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SEC Petition Evaluation Report 

Petition SEC-00224 

Report Rev Number: Rev. 0

Report Submittal Date: February 18, 2016

Subject Expert(s): Mitch Findley, Brian Gleckler, Jason Davis, Mike Mahathy

Site Expert(s): N/A

Petition Administrative Summary

Petition Under Evaluation

Petition Number: SEC-00224

Petition Type: 83.13

Petition Receipt Date: December 4, 2014

Qualification Date: March 13, 2015

DOE/AWE Facility Name: Argonne National Laboratory-West, Scoville, ID

Petition Class

Petitioner-Requested Class Definition: All workers who worked in any work location at the Argonne National 

Laboratory-West from January 1, 1949 through December 31, 1995, with the 

basis of inadequate monitoring for plutonium, neptunium and fission products 

during that time period.

Class Evaluated by NIOSH: All workers who worked in any location of the Argonne National LaboratoryWest from April 10, 1951 through December 31, 1979.

NIOSH-Proposed Class(es) to be Added 

to the SEC:

All employees of the Department of Energy, its predecessor agencies, and 

their contractors and subcontractors who worked at the Argonne National 

Laboratory-West between April 10, 1951 and December 31, 1957 for a 

number of work days aggregating at least 250 work days, occurring either 

solely under this employment, or in combination with work days within the 

parameters established for one or more other classes of employees in the 

Special Exposure Cohort.

Related Petition Summary Information

SEC Petition Tracking Number(s): N/A

Petition Type: N/A

DOE/AWE Facility Name: N/A

Petition Status: N/A

Related Evaluation Report Information

Report Title: N/A

DOE/AWE Facility Name: N/A

ORAU Preparation and Review

ORAU Lead Technical Evaluator: Mitch Findley

ORAU Peer Review Completed By: Dan Stempfley
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DCAS Review and Approval

Peer Review Completed By:

[Signature on File]

Timothy Taulbee 

February 18, 2016

SEC Petition Evaluation Reviewed By:

[Signature on File]

LaVon Rutherford

February 18, 2016

SEC Petition Evaluation Reviewed By:

[Signature on File]

Stuart L. Hinnefeld 

February 18, 2016 
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Evaluation Report Summary: SEC-00224, ANL-W 

This evaluation report by the National Institute for Occupational Safety and Health (NIOSH)

addresses a class of employees proposed for addition to the Special Exposure Cohort (SEC) per the 

Energy Employees Occupational Illness Compensation Program Act of 2000, as amended, 42 U.S.C. 

§ 7384 et seq. (EEOICPA) and 42 C.F.R. pt. 83, Procedures for Designating Classes of Employees as 

Members of the Special Exposure Cohort under the Energy Employees Occupational Illness 

Compensation Program Act of 2000. 

Petitioner-Requested Class Definition

Petition SEC-00224 was received on December 4, 2014, and qualified on March 13, 2015. The 

petitioner requested that NIOSH consider the following class: All workers who worked in any work 

location at the Argonne National Laboratory-West from January 1, 1949 through December 31, 1995, 

with the basis of inadequate monitoring for plutonium, neptunium and fission products during that 

time period. 

Class Evaluated by NIOSH

Based on its preliminary research, NIOSH modified the petitioner-requested class. NIOSH evaluated 

the following class: All workers who worked in any location of the Argonne National LaboratoryWest from April 10, 1951 through December 31, 1979. The petitioner-proposed class was modified 

(see Section 3.0 below). The start date was modified because the first reactor facility was completed, 

and its associated reactor assembly commenced, on April 10, 1951. The end date was modified 

because abundant plutonium bioassay data has been identified beginning in 1980. 

NIOSH-Proposed Class(es) to be Added to the SEC 

Based on its full research of the class under evaluation, NIOSH has defined a single class of 

employees for which NIOSH cannot estimate radiation doses with sufficient accuracy. The NIOSHproposed class includes all employees of the Department of Energy, its predecessor agencies, and their 

contractors and subcontractors who worked at the Argonne National Laboratory-West between April 

10, 1951 and December 31, 1957 for a number of work days aggregating at least 250 work days, 

occurring either solely under this employment, or in combination with work days within the 

parameters established for one or more other classes of employees in the Special Exposure Cohort. 

The class under evaluation was modified (see Section 3.0 below) because there are sufficient routine 

internal and external exposure data beginning on January 1, 1958 to complete dose reconstructions 

with sufficient accuracy.

Feasibility of Dose Reconstruction 

NIOSH finds it is not feasible to estimate internal and external exposures with sufficient accuracy for 

all workers at the ANL-W site from April 10, 1951 through December 31, 1957 due to insufficient 

data available during this time period. With the exception of this class, per EEOICPA and 42 C.F.R. § 

83.13(c)(1), NIOSH has established that it has access to sufficient information to: (1) estimate the 

maximum radiation dose, for every type of cancer for which radiation doses are reconstructed, that 

could have been incurred in plausible circumstances by any member of the class; or (2) estimate 
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radiation doses more precisely than an estimate of maximum dose. Information available from the site 

profile and additional resources is sufficient to document or estimate the maximum internal and 

external potential exposure to members of the evaluated class under plausible circumstances 

beginning January 1, 1958. 

The NIOSH dose reconstruction feasibility findings are based on the following: 

• The Argonne National Laboratory-West site, originally known as the Idaho Division of the 

Argonne National Laboratory or the Idaho Site, was established in 1949 and was operated as a 

separate site until 2005 when ANL-W was merged with the Idaho National Engineering and 

Environmental Laboratory (INEEL) to form the Idaho National Laboratory (INL). NIOSH 

determined that the most effective way of analyzing the ANL-W facilities was by focusing on the

two ANL-W operating areas on the INL Site: the EBR-I Complex and the EBR-II Complex. The 

EBR-I Complex was originally commissioned to test a fuel-breeding reactor; later operations 

involved building and testing experimental reactors. The EBR-II Complex included test reactor 

facilities and support facilities for reactor fuel experiments and testing.

• Principal sources of internal radiation for members of the evaluated ANL-W class working in the 

EBR-I and EBR-II Complexes included exposures to various isotopes of uranium, plutonium, 

thorium, and other actinides as well as exotic radionuclides (produced from, or as a result of, 

reactor neutron irradiation) that include mixed fission and activation products (MFP/MAP), 

radioiodines, and other radionuclides. 

• The internal dose potential during the period under evaluation was associated with reactors, 

laboratories, and other research and support activities as well as decontamination and 

decommissioning activities and radioactive waste disposal. There was limited personnel internal 

exposure potential to alpha-emitting radionuclides during the evaluated period. However, for the 

period from April 10, 1951 through December 31, 1957, NIOSH has not located sufficient 

personnel or area monitoring documentation to support complete reconstruction of internal 

personnel exposures. NIOSH has identified personnel monitoring data for mixed fission product 

exposures that are sufficient to support reconstructing internal doses for ANL-W workers 

beginning in 1958. 

• Principal sources of external radiation for members of the evaluated ANL-W class working in the 

EBR-I and EBR-II Complexes included beta-gamma and neutron exposures to various isotopes of 

uranium, plutonium, thorium, and other actinides as well as exotic radionuclides (produced from, 

or as a result of, reactor neutron irradiation) that include mixed fission and activation products 

(MFP/MAP), radioiodines, noble gases, and other radionuclides. By policy, ANL-W required 

external dosimetry (film or TLD) be worn within the operating (fenced) area. This dosimetry was 

issued at the security guard checkpoint. 
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• NIOSH has determined that the external dose potential during the period under evaluation was 

associated with reactors, laboratories, and other research and support activities as well as 

decontamination and decommissioning activities and radioactive waste disposal. For the period 

from April 10, 1951 through December 31, 1957, NIOSH has not located sufficient personnel or 

area monitoring documentation to support complete reconstruction of external personnel 

exposures. NIOSH has identified personnel monitoring data for external exposures that are 

sufficient to support reconstructing external doses for ANL-W workers beginning in 1958. 

• NIOSH finds that it is likely feasible to reconstruct occupational medical dose for ANL-W 

workers with sufficient accuracy.

• Pursuant to 42 C.F.R. § 83.13(c)(1), NIOSH determined that there is insufficient information for 

ANL-W personnel for the period from April 10, 1951 through December 31, 1957 to either: (1) 

estimate the maximum radiation dose, for every type of cancer for which radiation doses are 

reconstructed, that could have been incurred under plausible circumstances by any member of the 

class; or (2) estimate the radiation doses of members of the class more precisely than a maximum 

dose estimate.

• Although NIOSH found that it is not possible to completely reconstruct radiation doses for the 

proposed class, NIOSH intends to use any internal and external monitoring data that may become 

available for an individual claim (and that can be interpreted using existing NIOSH dose 

reconstruction processes or procedures). Therefore, dose reconstructions for individuals employed 

at ANL-W during the period from April 10, 1951 through December 31, 1957, but who do not 

qualify for inclusion in the SEC, may be performed using these data as appropriate.

Health Endangerment Determination

Per EEOICPA and 42 C.F.R. § 83.13(c)(3), a health endangerment determination is required because 

NIOSH has determined that it does not have sufficient information to estimate dose for the members 

of the proposed class from April 10, 1951 through December 31, 1957. 

NIOSH did not identify any evidence supplied by the petitioners or from other resources that would 

establish that the proposed class was exposed to radiation during a discrete incident likely to have 

involved exceptionally high-level exposures. However, evidence indicates that some workers in the 

proposed class may have accumulated substantial chronic exposures through episodic intakes of 

radionuclides, combined with external exposures to gamma, beta, and neutron radiation. 

Consequently, NIOSH has determined that health was endangered for those workers covered by this 

evaluation who were employed for at least 250 aggregated work days either solely under their 

employment or in combination with work days within the parameters established for other SEC 

classes. 

For the period January 1, 1958 through December 31, 1979, a health endangerment determination is 

not required because NIOSH has determined that it has sufficient information to estimate dose for the 

members of the evaluated class during this time frame. 
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Table of Contents

1.0 Purpose and Scope ........................................................................................................................ 15

2.0 Introduction .................................................................................................................................. 15

3.0 SEC-00224 Argonne National Laboratory-West Class Definitions ............................................. 16

3.1 Petitioner-Requested Class Definition and Basis ................................................................ 16

3.2 Class Evaluated by NIOSH ................................................................................................. 17

3.3 NIOSH-Proposed Class(es) to be Added to the SEC .......................................................... 17

4.0 Data Sources Reviewed by NIOSH to Evaluate the Class ........................................................... 17

4.1 Site Profile Technical Basis Documents (TBDs) ................................................................ 20

4.2 Technical Information Bulletins and Procedures ................................................................ 20

4.3 Facility Employees and Experts .......................................................................................... 22

4.4 Previous Dose Reconstructions ........................................................................................... 24

4.5 NIOSH Site Research Database .......................................................................................... 24

4.6 Documentation and/or Affidavits Provided by Petitioners ................................................. 24

5.0 Radiological Operations Relevant to the Class Evaluated by NIOSH ......................................... 25

5.1 Argonne National Laboratory-West Plant and Process Descriptions .................................. 25

5.1.1 EBR-I Complex ....................................................................................................... 27

5.1.1.1 Experimental Breeder Reactor-I (EBR-I) ................................................. 29

5.1.1.2 Zero Power Reactor-III (ZPR-III) ............................................................ 36

5.1.1.3 Boiling Water Reactor Experiments I-V (BORAX-I – BORAX-V) ........ 41

5.1.1.4 Argonne Fast Source Reactor (AFSR) (1st location: 1959-1970) ............ 51

5.1.2 EBR-II Complex ...................................................................................................... 54

5.1.2.1 Transient Reactor Experiment and Test Facility (TREAT)...................... 57

5.1.2.2 Experimental Breeder Reactor-II (EBR-II) .............................................. 62

5.1.2.3 Fuel Cycle Facility (FCF) ......................................................................... 68

5.1.2.4 Fuel Assembly & Storage Building (FASB) ............................................ 93

5.1.2.5 Inspection & Test Facility (ITF) ............................................................... 94

5.1.2.6 Hot Fuel Examination Facility-North (HFEF-N) ..................................... 95

5.1.2.7 Zero Power Plutonium Reactor (ZPPR) ................................................. 103

5.1.2.8 Laboratory & Operations Building (L&O Building) .............................. 108

5.1.2.9 Argonne Fast Source Reactor (AFSR) (2nd location): 1970-1997 ........ 112

5.1.2.10 Miscellaneous Facilities ......................................................................... 113

5.2 Internal Radiological Exposure Sources from ANL-W Site Operations ........................... 116

5.2.1 Internal Radiological Exposure Sources from EBR-I Complex Operations ......... 117

5.2.1.1 Fission and Activation Products ............................................................. 118

5.2.1.2 Radioiodines ........................................................................................... 118

5.2.1.3 Radioactive Noble Gases ........................................................................ 118

5.2.1.4 Uranium .................................................................................................. 118

5.2.1.5 Plutonium ............................................................................................... 119

5.2.1.6 Thorium .................................................................................................. 120

5.2.2 Internal Radiological Exposure Sources from EBR-II Complex Operations ........ 120
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5.2.2.1 Fission and Activation Products ............................................................. 121

5.2.2.2 Radioiodines ........................................................................................... 121

5.2.2.3 Radioactive Noble Gases ........................................................................ 121

5.2.2.4 Plutonium ............................................................................................... 122

5.2.2.5 Uranium .................................................................................................. 123

5.2.2.6 Thorium .................................................................................................. 124

5.2.2.7 Other Actinides ....................................................................................... 125

5.2.2.8 Other Radionuclides ............................................................................... 125

5.3 External Radiological Exposure Sources from ANL-W Site Operations ......................... 126

5.3.1 External Radiological Exposure Sources from EBR-I Complex Operations ........ 128

5.3.1.1 Photon ..................................................................................................... 128

5.3.1.2 Beta ......................................................................................................... 129

5.3.1.3 Neutron ................................................................................................... 130

5.3.2 External Radiological Exposure Sources from EBR-II Complex Operations ....... 131

5.3.2.1 Photon ..................................................................................................... 131

5.3.2.2 Beta ......................................................................................................... 133

5.3.2.3 Neutron ................................................................................................... 134

5.4 Incidents at the ANL-W Site ............................................................................................. 135

5.4.1 Incidents at the EBR-I Complex ............................................................................ 135

5.4.1.1 Excursion at BORAX-I (July 1954) ....................................................... 135

5.4.1.2 Suspected Brick Rupture in EBR-I Breeding Blanket (April 1955) ...... 139

5.4.1.3 Partial Meltdown of EBR-I Core (November 1955) .............................. 139

5.4.1.4 Personnel Contamination at BORAX-IV (July 1957) ............................ 139

5.4.1.5 Airborne Contamination at BORAX-IV (September 1957) ................... 139

5.4.1.6 Fuel Defect Test at BORAX-IV (March 1958) ...................................... 140

5.4.1.7 Fire in EBR-I Rod Wash Room (September 1959) ................................ 140

5.4.1.8 Contamination on Plutonium Plates at ZPR-III (March 1960) .............. 140

5.4.1.9 Alpha Contamination at AFSR (November 1963) ................................. 141

5.4.1.10 Defective Fuel Test at BORAX-V (August 1964) ................................. 141

5.4.1.11 Spontaneous Combustion of U-235 Foils at ZPR-III (November 1966) 141

5.4.1.12 Leaking Po-Be Source at ZPR-III (April 1967) ..................................... 141

5.4.2 Incidents at the EBR-II Complex .......................................................................... 142

5.4.2.1 Perchloric Acid Explosion in L&O Chemistry Lab 40 (July 1962) ....... 142

5.4.2.2 L&O Building Broken Specimen (March 1963) .................................... 142

5.4.2.3 Hydrogen Explosion in EBR-II/FCF Transfer Lock (Dec. 1965) .......... 142

5.4.2.4 EBR-II Reactor Radioactive Sodium Fire (June 1967) .......................... 143

5.4.2.5 FCF Contaminated Water Spill (June 1968) .......................................... 143

5.4.2.6 ZPPR-AFSR Dropped Blanket Drawer (January 1975) ......................... 143

6.0 Summary of Available Monitoring Data for the Class Evaluated by NIOSH ........................... 144

6.1 Available ANL-W Internal Monitoring Data .................................................................... 144

6.2 Available ANL-W External Monitoring Data ................................................................... 148

6.3 Available ANL-W Radiological Monitoring Data ............................................................ 153

6.3.1 Available ANL-W Radiological Monitoring Data for EBR-I Complex ............... 155

6.3.1.1 Available Radiological Monitoring Data for EBR-I .............................. 155

6.3.1.2 Available Radiological Monitoring Data for ZPR-III ............................ 157

6.3.1.3 Available Radiological Monitoring Data for BORAX ........................... 159
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6.3.1.4 Available Radiological Monitoring Data for AFSR ............................... 163

6.3.2 Available ANL-W Radiological Monitoring Data for EBR-II Complex .............. 166

6.3.2.1 Available Radiological Monitoring Data for TREAT ............................ 166

6.3.2.2 Available Radiological Monitoring Data for EBR-II Reactor ............... 169

6.3.2.3 Available Radiological Monitoring Data for FCF .................................. 172

6.3.2.4 Available Radiological Monitoring Data for FASB ............................... 175

6.3.2.5 Available Radiological Monitoring Data for ITF ................................... 178

6.3.2.6 Available Radiological Monitoring Data for HFEF-N ........................... 181

6.3.2.7 Available Radiological Monitoring Data for ZPPR ............................... 184

6.3.2.8 Available Radiological Monitoring Data for the L&O Building ........... 185

7.0 Feasibility of Dose Reconstruction for the Class Evaluated by NIOSH .................................... 188

7.1 Pedigree of ANL-W Data .................................................................................................. 188

7.1.1 Internal Monitoring Data Pedigree Review ........................................................... 189

7.1.2 External Monitoring Data Pedigree Review .......................................................... 189

7.1.3 Area Monitoring Data Pedigree Review ............................................................... 190

7.2 Evaluation of Bounding Internal Radiation Doses at ANL-W .......................................... 190

7.2.1 Evaluation of Bounding Internal Radiation Doses for the EBR-I Complex ......... 191

7.2.1.1 Evaluation of Bounding Process-Related Internal Doses for the EBR-I 

Complex ................................................................................................. 192

7.2.1.2 Methods for Bounding Operational Period Internal Doses for the EBR-I 

Complex ................................................................................................. 196

7.2.2 Evaluation of Bounding Internal Radiation Doses for the EBR-II Complex ........ 198

7.2.2.1 Evaluation of Bounding Process-Related Internal Doses for the EBR-II 

Complex ................................................................................................. 198

7.2.2.2 Methods for Bounding Operational Period Internal Doses for the EBR-II 

Complex ................................................................................................. 199

7.2.3 Internal Dose Reconstruction Feasibility Conclusion ........................................... 208

7.3 Evaluation of Bounding External Radiation Doses at ANL-W ......................................... 209

7.3.1 Evaluation of Bounding Process-Related External Doses ..................................... 209

7.3.2 Evaluation of Bounding Ambient Environmental External Doses ........................ 211

7.3.3 ANL-W Occupational X-Ray Examinations ......................................................... 211

7.3.4 Methods for Bounding External Dose at ANL-W ................................................. 212

7.3.4.1 Methods for Bounding Operational Period External Dose ..................... 212

7.3.4.2 Methods for Bounding Ambient Environmental External Doses ........... 212

7.3.5 External Dose Reconstruction Feasibility Conclusion .......................................... 213

7.4 Evaluation of Petition Basis for SEC-00224 ..................................................................... 214

7.5 Other Potential SEC Issues Relevant to the Petition Identified During the Evaluation .... 214

7.6 Summary of Feasibility Findings for Petition SEC-00224 ................................................ 215
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8.0 Evaluation of Health Endangerment for Petition SEC-00224 .................................................... 216

9.0 Class Conclusion for Petition SEC-00224 ................................................................................. 217

10.0 References .................................................................................................................................. 219

Attachment 1: Data Capture Synopsis ................................................................................................. 245

Attachment 2: Acronyms and Abbreviations ...................................................................................... 257

Tables 

Table 4-1: Data Captures Conducted in Support of the SEC-00224 ANL-W Evaluation .................... 19

Table 4-2: No. of ANL-W Claims Submitted Under the Dose Reconstruction Rule ........................... 24

Table 5-1: BORAX Reactors Operational Periods ................................................................................ 41

Table 5-2: EBR-II Timeline (1957-1979) ............................................................................................. 67

Table 5-3: Important Dates in FCF Operating History ......................................................................... 90

Table 5-4: INL Site Beta-Gamma Dosimetry Systems History and LOD Summary .......................... 127

Table 5-5: INL Site Neutron Dosimetry Systems History and LOD Summary .................................. 128

Table 6-1: Bioassay Results for 50 Selected ANL-W Workers .......................................................... 145

Table 6-2: Trip Settings for ZPPR Particulate Air Monitors .............................................................. 184

Table 7-1: Internal Monitoring Data from All Sources for 12 Selected Workers (1953-1960) .......... 193

Table 7-2: Internal Monitoring Data Provided by Sites for 12 Selected Workers (1953-1960) ......... 194

Table 7-3: Monitoring of AFSR Workers ........................................................................................... 195

Table 7-4: Monitoring of BORAX-V Workers ................................................................................... 195

Table 7-5: Default Ratios for Recycled Uranium Intakes .................................................................. 201

Table 7-6: Available External Monitoring Data for 50 Selected ANL-W Workers ........................... 209

Table 7-7: Available External Monitoring Data for 12 Selected Workers (1952-1960) ..................... 210

Table 7-8: Summary of Feasibility Findings for SEC-00224 ............................................................. 215

Figures 

Figure 5-1: Locations of ANL-W Facilities (EBR-I Complex and EBR-II Complex) ......................... 26

Figure 5-2: Aerial View of the Main Part of the EBR-I Complex ........................................................ 27

Figure 5-3: Aerial View of the Entire EBR-I Complex ........................................................................ 28

Figure 5-4: Installation of EBR-I Reactor Vessel ................................................................................. 29

Figure 5-5: Four Light Bulbs Illuminated by EBR-1 (Dec. 20, 1951) .................................................. 30

Figure 5-6: Cut-away View of EBR-I ................................................................................................... 31

Figure 5-7: View of EBR-I .................................................................................................................... 32

Figure 5-8: EBR-I Reactor Systems ...................................................................................................... 33

Figure 5-9: View of Melted Mark II Core Assembly ............................................................................ 35

Figure 5-10: ZPR-III Floor Plan ............................................................................................................ 37
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Figure 5-11: ZPR-III Critical Assembly Machine ................................................................................. 38

Figure 5-12: Example of a ZPR-III Core Loading ................................................................................ 39

Figure 5-13: Typical Loaded ZPR-III Fuel Drawer .............................................................................. 40

Figure 5-14: BORAX-I Site Plan .......................................................................................................... 42

Figure 5-15: BORAX-I, Vertical Diagram ............................................................................................ 43

Figure 5-16: Steam from Excursion of BORAX-I ................................................................................ 44

Figure 5-17: BORAX II ......................................................................................................................... 46

Figure 5-18: BORAX-III Turbine ......................................................................................................... 47

Figure 5-19: BORAX-III Facility .......................................................................................................... 48

Figure 5-20: BORAX-V Facility ........................................................................................................... 50

Figure 5-21: AFSR Building Layout (1st location) ................................................................................ 52

Figure 5-22: AFSR Building (1st location) ............................................................................................ 53

Figure 5-23: Aerial View of the EBR-II Complex (circa early-1960s) ................................................. 55

Figure 5-24: Map of EBR-II Complex in 1981 ..................................................................................... 56

Figure 5-25: TREAT Reactor Building Plan ......................................................................................... 58

Figure 5-26: TREAT Reactor Plan View .............................................................................................. 60

Figure 5-27: EBR-II Reactor Building (side view) ............................................................................... 63

Figure 5-28: EBR-II Reactor Building (top view by levels) ................................................................. 64

Figure 5-29: The EBR-II Reactor’s Four Operating Missions .............................................................. 65

Figure 5-30: General Layout of FCF’s Main Floor (ground floor) ....................................................... 70

Figure 5-31: Original Layout of FCF’s Basement ................................................................................ 71

Figure 5-32: Elevation Cross-Section of FCF ....................................................................................... 72

Figure 5-33: Elevation Cross Section of Reactor Passageway .............................................................. 75

Figure 5-34: New Hot Repair Area in FCF’s Basement ....................................................................... 80

Figure 5-35: Approximate Distribution of Elements in Melt-Refining Process ................................... 83

Figure 5-36: Gaseous Effluent Treatment System for FCF Air and Argon Cells ................................. 87

Figure 5-37: General Layout of FASB .................................................................................................. 93

Figure 5-38: General Layout of ITF ...................................................................................................... 94

Figure 5-39: Elevation Cross-Sections of HFEF-N ............................................................................... 96

Figure 5-40: Floor Plan of HFEF-N Operating Floor ........................................................................... 97

Figure 5-41: Elevation Detail of the NRAD Reactor and Radiography Room at HFEF-N .................. 98

Figure 5-42: ZPPR Reactor and Support Areas .................................................................................. 104

Figure 5-43: ZPPR Vault Workroom Area ......................................................................................... 105

Figure 5-44: The Analytical Laboratory Wing of the L&O Building ................................................. 109

Figure 5-45: The Analytical Laboratory Junior Caves ........................................................................ 110

Figure 5-46: ANL-W Radioactive Scrap Waste Facility (foreground) ............................................... 114

Figure 5-47: RSWF Storage Liner Array and Schematic .................................................................... 115

Figure 5-48: Destructive Test of BORAX-I ........................................................................................ 136

Figure 5-49: Spectra Comparison: BORAX-I and Pacific Proving Grounds ...................................... 138

Figure 6-1: Bioassay Monitoring Data Trend ..................................................................................... 146

Figure 6-2: Total Internal Monitoring Performed at ANL-W (1956-1979) ........................................ 147

Figure 6-3: External Monitoring Data Trend ...................................................................................... 149

Figure 6-4: ANL-W Badges Issued by Area (1952-1966) .................................................................. 151

Figure 6-5: Number of Persons Monitored at ANL-W by Area (1958-1974) .................................... 152

Figure 6-6: Example of Air Sampling Data Sheet at EBR-I in 1955 .................................................. 156

Figure 6-7: Example of Alpha Contamination Smear Survey at EBR-I in 1961 ................................ 157
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Figure 6-8: Example of Check Sheet Used for Plutonium Loading .................................................... 158

Figure 6-9: Example of “Piece Counting” of Plutonium Fuel Plates .................................................. 159

Figure 6-10: Example of Radiation Survey of BORAX-V Reactor Main Floor ................................. 160

Figure 6-11: Example of Smear Survey of BORAX Reactor Main Floor .......................................... 161

Figure 6-12: Example of Air Monitoring Results for BORAX-V Turbine Building .......................... 162

Figure 6-13: Example of AFSR Radiation Survey with Smears ......................................................... 164

Figure 6-14: Example of ASFR Neutron Survey ................................................................................ 165

Figure 6-15: Example of TREAT Reactor Room Radiation Survey with Smears .............................. 167

Figure 6-16: TREAT Radiation Survey Report Corresponding to Figure 6-15 .................................. 168

Figure 6-17: Example of EBR-II Reactor Survey Report from 1969 ................................................. 169

Figure 6-18: Example of EBR-II Reactor Smear Sample Data Sheet from 1969 ............................... 170

Figure 6-19: Example of EBR-II Reactor Air Sample Data Sheet from 1969 .................................... 171

Figure 6-20: Example of FCF Survey Report from 1965 .................................................................... 172

Figure 6-21: Example of FCF Air Sample Data Sheet from 1965 ...................................................... 173

Figure 6-22: Example of FCF Smear Sample Data Sheet from 1965 ................................................. 174

Figure 6-23: Example of FASB Survey Report from 1978 ................................................................. 175

Figure 6-24: Example of FASB Air Sample Data Sheet from 1978 ................................................... 176

Figure 6-25: Example of FASB Smear Sample Data Sheet from 1973 .............................................. 177

Figure 6-26: Example of ITF Survey Report from 1969 ..................................................................... 178

Figure 6-27: Example of ITF Air Sample Data Sheet from 1969 ....................................................... 179

Figure 6-28: Example of ITF Smear Sample Data Sheet from 1969 .................................................. 180

Figure 6-29: Example of HFEF-N Survey Report from 1977 ............................................................. 181

Figure 6-30: Example of HFEF-N Air Sample Data Sheet from 1976 ............................................... 182

Figure 6-31: Example of HFEF-N Smear Sample Data Sheet from 1976 .......................................... 183

Figure 6-32: Example ZPPR Smear Sample Data Sheet from 1973 ................................................... 185

Figure 6-33: Example L&O Building Routine Surveys – 1973 .......................................................... 186

Figure 6-34: Example L&O Building Junior Cave Area Survey Map ................................................ 187

Figure 7-1: Example FCF Cold-Line Air Sample Result from 1969 .................................................. 203

Figure 7-2: Example FCF Room 25 Air Sample Result from 1963 .................................................... 205

Figure 7-3: Example Air Sample Result for FCF RAS-TREAT Sodium-Loop Work from 1970 ...... 207
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Acronyms and Abbreviations

An acronyms and abbreviations list is available in Attachment 2 at the end of this document.
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SEC Petition Evaluation Report for SEC-00224

1.0 Purpose and Scope

This report evaluates the feasibility of reconstructing doses for all workers who worked in any 

location of the Argonne National Laboratory-West from April 10, 1951 through December 31, 1979. 

It provides information and analyses germane to considering a petition for adding a class of 

employees to the congressionally-created SEC. 

This report does not make any determinations concerning the feasibility of dose reconstruction that 

necessarily apply to any individual energy employee who might require a dose reconstruction from 

NIOSH. This report also does not contain the final determination as to whether the proposed class 

will be added to the SEC (see Section 2.0). 

This evaluation was conducted in accordance with the requirements of EEOICPA, 42 C.F.R. pt. 83, 

and the guidance contained in the Division of Compensation Analysis and Support’s (DCAS) Internal 

Procedures for the Evaluation of Special Exposure Cohort Petitions, DCAS-PR-004.1

1 DCAS was formerly known as the Office of Compensation Analysis and Support (OCAS).

2.0 Introduction 

Both EEOICPA and 42 C.F.R. pt. 83 require NIOSH to evaluate qualified petitions requesting that the 

Department of Health and Human Services (HHS) add a class of employees to the SEC. The 

evaluation is intended to provide a fair, science-based determination of whether it is feasible to 

estimate with sufficient accuracy the radiation doses of the class of employees through NIOSH dose 

reconstructions.2

2 NIOSH dose reconstructions under EEOICPA are performed using the methods promulgated under 42 C.F.R. pt. 82 and 

the detailed implementation guidelines available at http://www.cdc.gov/niosh/ocas.

42 C.F.R. § 83.13(c)(1) states: Radiation doses can be estimated with sufficient accuracy if NIOSH 

has established that it has access to sufficient information to estimate the maximum radiation dose, 

for every type of cancer for which radiation doses are reconstructed, that could have been incurred in 

plausible circumstances by any member of the class, or if NIOSH has established that it has access to 

sufficient information to estimate the radiation doses of members of the class more precisely than an 

estimate of the maximum radiation dose.

Under 42 C.F.R. § 83.13(c)(3), if it is not feasible to estimate with sufficient accuracy radiation doses 

for members of the class, then NIOSH must determine that there is a reasonable likelihood that such 

radiation doses may have endangered the health of members of the class. The regulation requires 

NIOSH to assume that any duration of unprotected exposure may have endangered the health of 

members of a class when it has been established that the class may have been exposed to radiation 

during a discrete incident likely to have involved levels of exposure similarly high to those occurring 

during nuclear criticality incidents. If the occurrence of such an exceptionally high-level exposure has 

not been established, then NIOSH is required to specify that health was endangered for those workers 

 
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who were employed for at least 250 aggregated work days within the parameters established for the 

class or in combination with work days within the parameters established for one or more other SEC 

classes. 

NIOSH is required to document its evaluation in a report, and to do so, relies upon both its own dose 

reconstruction expertise as well as technical support from its contractor, Oak Ridge Associated 

Universities (ORAU). Once completed, NIOSH provides the report to both the petitioner(s) and the 

Advisory Board on Radiation and Worker Health (Board). The Board will consider the NIOSH 

evaluation report, together with the petition, petitioner(s) comments, and other information the Board 

considers appropriate, in order to make recommendations to the Secretary of HHS on whether or not 

to add one or more classes of employees to the SEC. Once NIOSH has received and considered the 

advice of the Board, the Director of NIOSH will propose a decision on behalf of HHS. The Secretary 

of HHS will make the final decision, taking into account the NIOSH evaluation, the advice of the 

Board, and the proposed decision issued by NIOSH. As part of this decision process, petitioners may 

seek a review of certain types of final decisions issued by the Secretary of HHS.

3

3 See 42 C.F.R. pt. 83 for a full description of the procedures summarized here. Additional internal procedures are 

available at http://www.cdc.gov/niosh/ocas.

3.0 SEC-00224 Argonne National Laboratory-West Class Definitions 

The following subsections address the evolution of the class definition for SEC-00224, Argonne 

National Laboratory-West (ANL-W). When a petition is submitted, the requested class definition is 

reviewed as submitted. Based on its review of the available site information and data, NIOSH will 

make a determination whether to qualify for full evaluation all, some, or no part of the petitionerrequested class. If some portion of the petitioner-requested class is qualified, NIOSH will specify that 

class along with a justification for any modification of the petitioner’s class. After a full evaluation of 

the qualified class, NIOSH will determine whether to propose a class for addition to the SEC and will 

specify that proposed class definition. 

3.1 Petitioner-Requested Class Definition and Basis

Petition SEC-00224 was received on December 4, 2014 and qualified on March 13, 2015. The 

petitioner requested that NIOSH consider the following class: All workers who worked in any work 

location at the Argonne National Laboratory-West from January 1, 1949 through December 31, 1995, 

with the basis of inadequate monitoring for plutonium, neptunium and fission products during that 

time period. 

The petitioner provided information and affidavit statements in support of the petitioner’s belief that 

accurate dose reconstruction over time is impossible for the ANL-W workers in question. NIOSH 

deemed the following information and affidavit statements sufficient to qualify SEC-00224 for 

evaluation: 

The petitioner indicated in Section F of the Special Exposure Cohort Petition – Form B, that 

there was inadequate monitoring for plutonium, neptunium, and fission products (DSA Ref 

 
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ID: 120480). This serves as the basis for proposing that records and information at ANL-W 

were inadequate for individual dose reconstruction. 

Based on its ANL-W research and data capture efforts, NIOSH determined that it has access to 

personnel bioassay monitoring data (including urinalysis, fecal sampling, and whole-body counts), 

and external dosimetry data for ANL-W workers during the time period under evaluation. However, 

NIOSH also determined that internal and external monitoring records are not complete for all time 

periods or for all radionuclides. NIOSH concluded that there is sufficient documentation to support, 

for at least part of the requested time period, the petition basis that internal and external radiation 

exposures and radiation doses were not adequately monitored at ANL-W, either through personal 

monitoring or area monitoring. The information and statements provided by the petitioner qualified 

the petition for further consideration by NIOSH, the Board, and HHS. The details of the petition basis 

are addressed in Section 7.4. 

3.2 Class Evaluated by NIOSH

Based on its preliminary research, NIOSH modified the petitioner-requested class. The class start date 

was modified because ANL-W’s first radiologically-related operation, at EBR-I, did not occur until 

April 10, 1951, when the reactor facility was completed and its associated reactor assembly 

commenced. The class end date was modified because of the large number of plutonium bioassay 

analyses available beginning in 1980. Therefore, NIOSH defined the following class for further 

evaluation: All workers who worked in any location of the Argonne National Laboratory-West from 

April 10, 1951 through December 31, 1979. 

3.3 NIOSH-Proposed Class(es) to be Added to the SEC

Based on its research of the class under evaluation, NIOSH has defined a single class of employees for 

which NIOSH cannot estimate radiation doses with sufficient accuracy. The NIOSH-proposed SEC 

class includes all employees of the Department of Energy, its predecessor agencies, and their 

contractors and subcontractors who worked at the Argonne National Laboratory-West between April 

10, 1951 and December 31, 1957 for a number of work days aggregating at least 250 work days, 

occurring either solely under this employment, or in combination with work days within the 

parameters established for one or more other classes of employees in the Special Exposure Cohort. 

4.0 Data Sources Reviewed by NIOSH to Evaluate the Class 

Since Argonne National Laboratory-West was made part of Idaho National Laboratory in 2005, 

records generated by ANL-W have been integrated into INL’s records collection. INL uses the 

Electronic Document Management System (EDMS) to manage records that are in an electronic 

format. EDMS is made of up of multiple client/server and web applications that run on several 

different operating systems and hardware to provide thousands of users with electronic documents. 

Preset as well as user-defined queries can be performed. The NIOSH and ORAUT Leads were 

provided EDMS accounts to allow for searches of non-classified documents during data capture trips 

in Idaho. EDMS was the primary resource in identifying and capturing documents of interest for the 

SEC-00224 Evaluation Report. 
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As is standard practice, NIOSH also completed an extensive database and Internet search for 

information regarding ANL-W. The database search included the DOE Legacy Management 

Considered Sites database, the DOE Office of Scientific and Technical Information (OSTI) SciTech 

Connect database, the OSTI Energy Citations database, and the Hanford Declassified Document 

Retrieval System. In addition to general Internet searches, the NIOSH Internet search included OSTI 

OpenNet Advanced searches, OSTI Information Bridge Fielded searches, Nuclear Regulatory 

Commission (NRC) Agency-wide Documents Access and Management (ADAMS) web searches, the 

DOE-National Nuclear Security Administration-Nevada Site Office-search, the Defense Technical 

Information Center database (DTIC), the DOE website, the DOE Comprehensive Epidemiologic Data 

Resource (CEDR), the DOE Environmental Management website, the DOE Oak Ridge Operations 

website, the Energy Employees Claimant Assistance Project (EECAP) website, the Hathitrust website, 

the Internet edition of the Health Physics Journal, the National Academies Press website, the Oak 

Ridge National Laboratory online library and report repository, the University of Hawaii, Manoa 

online collection, the University of North Texas online collection, the U.S. Army Corps of Engineers 

website, and the U.S. Transuranium and Uranium website. Attachment 1 contains a summary of 

ANL-W documents. The summary specifically identifies data capture details and general descriptions 

of the documents retrieved. 

NIOSH and ORAUT staff went on five data capture trips to assess and scan ANL-W-related 

documentation at INL. These five trips resulted in the capture of a wide variety of relevant 

documentation, including 1,716 documents and 130,661 pages. In addition, ORAUT staff went on 

two data capture trips to Argonne National Laboratory-East (Lemont, IL), primarily to understand the 

structure of the employee occupational exposure histories held there, how these exposure history 

collections are accessed by ANL-E staff when responding to claimant dosimetry requests under 

EEOICPA, and to retrieve exposure history information on a selected group of early ANL-W workers. 

These two trips and follow-up inquiries resulted in the capture of exposure histories of 61 early 

ANL-W workers. The data captures at ANL-E also resulted in the collection of 92 documents and 

4,993 pages of relevant information. Table 4-1 lists the ANL-W data capture trips and the types of 

information obtained. 

In addition to the database and Internet searches listed above, NIOSH identified and reviewed 

numerous data sources to determine information relevant to determining the feasibility of dose 

reconstruction for the class of employees under evaluation. This included determining the availability 

of information on personal monitoring, area monitoring, industrial processes, and radiation source 

materials. The following subsections summarize the data sources identified and reviewed by NIOSH.
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Table 4-1: Data Captures Conducted in Support of the SEC-00224 ANL-W Evaluation

Site/Facility 

of Visit

Dates of 

Visit

Boxes 

Reviewed 

Docs 

Captured 

Page 

Count Brief Summary Of Site Visit

Idaho 

National 

Laboratory

09/21/2015 -

09/25/2015

59 95 14,527 Air sample data, radiological data, whole body counts, urinalysis data, and personnel dosimetry data

Idaho 

National 

Laboratory

04/27/2015 -

04/30/2015

48 652 40,481 Air sampling data, radiation surveys, daily log books, air monitoring, radiation reports, routine survey 

reports, smear data, technical specifications, radiation safety log books, personnel skin contamination, 

health physics log books, (Argonne National Laboratory); air samples, alpha contamination surveys, 

neutron surveys, safety analysis reports, early waste retrieval, offsite incoming radioactive material 

shipment records, health physics log books (Idaho National Laboratory).

Idaho 

National 

Laboratory

04/13/2015 -

04/16/2015

107 318 37,181 Visitor cards, individual worker's medical files, incidents, film badge reports, air and smear reports, 

temporary badge reports, radiological surveys, quarterly exposure summaries, bioassay results, 

contamination surveys, safety training records, health physics weekly reports, urine sample results, 

personnel exposure questionnaires, appraisal documents, and skin contamination results.

Idaho 

National 

Laboratory

03/29/2015 -

04/02/2015

91 489 26,731 Controlled entry procedure, shipping incidents and related contamination, hazards analysis, photographic 

history of the Argonaut Multiplication Experiment, locating and identifying the source of the 5/24/67 

fission product release in EBR-II, air sample data, radiological surveys, sampling and smear data, 

external dose reports, emergency preparedness memos, inspection reports, gas cooler reactor experiment 

information, pending whole body count data to be cross-checked against existing data sets, repair of 

central superheater assemblies, contamination of EBR-II, EBR-II reports, safety procedures, actinide fuel 

fabrication, health physics logbook, and EBR-II run reports.

Idaho 

National 

Laboratory

12/08/2008 -

12/12/2008

62 70 6,748 Monthly HP reports (1968 - 1969), neutron surveys, one month sample of radiological surveys, 

personnel exposure information (1953 - 1961), routine urinalysis (1963 - 1966), routine whole body 

counts (1964 - 1971), radiological incidents, air sample data sheets, employee exposure information, 

whole body and hand exposure (1969 - 19 71). 

Argonne 

National 

Laboratory - 

East

10/12/2015 -

10/14/2015

None None 0 Reconnaissance trip to better understand the structure of the employee occupational exposure histories 

held there, how these exposure history collections are accessed by Argonne National Laboratory-East 

staff when responding to claimant dosimetry requests under EEOICPA, and to retrieve exposure history 

information on a selected group of early Argonne National Laboratory-West workers.

Argonne 

National 

Laboratory - 

East

06/15/2015 -

06/19/2015

82 92 4,993 Radiation safety section monthly reports, air sample data, explosion report at EBR-II, technical 

specifications for EBR-II, health physic reports, bioassay reports, hazard summary report for EBR-II and 

addendum, occupational safety and health audit of EBR-II, Idaho reactor division personnel list, safety 

analysis of the EBR-II reprocessing facility, and tritium in EBR-II.

TOTAL N/A 449 1,716 130,661 N/A
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4.1 Site Profile Technical Basis Documents (TBDs) 

A Site Profile provides specific information concerning the documentation of historical practices at 

the specified site. Dose reconstructors can use the Site Profile to evaluate internal and external 

dosimetry data for monitored and unmonitored workers, and to supplement, or substitute for, 

individual monitoring data. A Site Profile consists of an Introduction and five Technical Basis 

Documents (TBDs) that provide process history information, information on personal and area 

monitoring, radiation source descriptions, and references to primary documents relevant to the 

radiological operations at the site. The Site Profile for a small site may consist of a single document. 

As part of NIOSH’s evaluation detailed herein, it examined the following TBDs for insights into 

ANL-W operations or related topics/operations at other sites: 

• TBD for INL - ANL-W – Introduction, ORAUT-TKBS-0007-1; Rev. 03; March 12, 2010; SRDB 

Ref ID: 79862 

• TBD for INL – ANL-W – Site Description, ORAUT-TKBS-0007-2; Rev. 04; August 2, 2010; 

SRDB Ref ID: 84106 

• TBD for INL – ANL-W – Occupational Medical Dose, ORAUT-TKBS-0007-3; Rev. 02 ; 

December 21, 2009; SRDB Ref ID: 77829 

• TBD for INL – ANL-W – Occupational Environmental Dose, ORAUT-TKBS-0007-4; Rev. 02; 

January 8, 2010; SRDB Ref ID: 78635 

• TBD for INL – ANL-W – Occupational Internal Dose, ORAUT-TKBS-0007-5; Rev. 03; March 2, 

2010; SRDB Ref ID: 79571 

• TBD for INL – ANL-W – Occupational External Dose, ORAUT-TKBS-0007-6; Rev. 03; April 19, 

2011; SRDB Ref ID: 94104 

• TBD for Y-12 National Security Complex – Occupational Internal Dose, ORAUT-TKBS-0014-5, 

Rev. 03; Oak Ridge Associated Universities; March 12, 2012; SRDB Ref ID: 109202 

4.2 Technical Information Bulletins and Procedures

A Technical Information Bulletin is a general ORAUT or NIOSH working document that provides 

guidance for preparing dose reconstructions at particular sites or categories of sites. An ORAUT 

Procedure provides specific requirements and guidance regarding EEOICPA project-level activities, 

including preparation of dose reconstructions at particular sites or categories of sites. NIOSH 

reviewed the following procedure as part of its evaluation: 

• OTIB: Internal Dosimetry Organ, External Dosimetry Organ, and IREP Model Selection by ICD9 Code, ORAUT-OTIB-0005 Rev 05; December 20, 2012; SRDB Ref ID: 121336 

• OTIB: Technical Information Bulletin: Tritium Calculated and Missed Dose Estimates, ORAUTOTIB-0011 Rev 00; June 29, 2004; SRDB Ref ID: 19430 
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• OTIB: Assignment of Environmental Internal Doses for Employees Not Exposed to Airborne 

Radionuclides in the Workplace, ORAUT-OTIB-0014 Rev 00; June 22, 2004; SRDB Ref ID: 

19432 

• OTIB: Bulletin: Interpretation of Dosimetry Data for Assignment of Shallow Dose, ORAUT-OTIB0017 Rev 01; October 11, 2005; SRDB Ref ID: 19434 

• OTIB: Internal Dose Overestimates for Facilities with Air Sampling Programs, ORAUT-OTIB0018 Rev 01; August 9, 2005; SRDB Ref ID: 19436 

• OTIB: Technical Information Bulletin: Assignment of Missed Neutron Doses Based on Dosimeter 

Records, ORAUT-OTIB-0023 Rev 01; May 14, 2008; SRDB Ref ID: 43937 

• OTIB: Application of Internal Doses Based on Claimant-Favorable Assumptions for Processing 

as Best Estimates, ORAUT-OTIB-0033 Rev 00; April 20, 2005; SRDB Ref ID: 19457 

• OTIB: Estimating Doses for Plutonium Strongly Retained in the Lung, ORAUT-OTIB-0049 Rev 01 

PC-2; November 29, 2010; SRDB Ref ID: 90666 

• OTIB: Technical Information Bulletin: Fission and Activation Product Assignment for Internal 

Dose-Related Gross Beta and Gross Gamma Analyses, ORAUT-OTIB-0054 Rev 02; March 6, 

2014; SRDB Ref ID: 130852 

• OTIB: Technical Information Bulletin: Internal Dose Reconstruction, ORAUT-OTIB-0060 Rev 00; 

February 6, 2007; SRDB Ref ID: 29984 

• OTIB: Guidance on Assigning Occupational X-Ray Dose under EEOICPA for X-Rays 

Administered Off Site, ORAUT-OTIB-0079 Rev 00; January 3, 2011; SRDB Ref ID: 89563 

• OTIB: Technical Information Bulletin: Dose Reconstruction Method for Chronic Lymphocytic 

Leukemia, ORAUT-OTIB-0082 Rev 00 PC-1; December 20, 2012; SRDB Ref ID: 121337 

• TIB: Estimations of Ingestion Intakes, OCAS-TIB-009; Rev. 00; April 13, 2004; SRDB Ref ID: 

22397 

• Procedure: Occupational On-Site Ambient Dose Reconstruction for DOE Sites, ORAUT-PROC0060 Rev 01; June 28, 2006; SRDB Ref ID: 29986 

• Procedure: Occupational Medical X-Ray Dose Reconstruction for DOE Sites, Rev 03, ORAUTPROC-0061 Rev 03; March 3, 2010; SRDB Ref ID: 79758 
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4.3 Facility Employees and Experts

To obtain additional information, NIOSH interviewed 19 former and current ANL-W employees. 

• Personal Communication, 2015a, Personal Communication with a former ANL-W chemist; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 18, 2015; SRDB Ref ID: 

147021 

• Personal Communication, 2015b, Personal Communication with a former ANL-W engineer; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 19, 2015; SRDB Ref ID: 

147018 

• Personal Communication, 2015c, Personal Communication with a former ANL-W auditor; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 19, 2015; SRDB Ref ID: 

147248 

• Personal Communication, 2015d, Personal Communication with a former ANL-W operator; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 19, 2015; SRDB Ref ID:

147009 

• Personal Communication, 2015e, Personal Communication with a former ANL-W manager; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 19, 2015; SRDB Ref ID: 

147010 

• Personal Communication, 2015f, Personal Communication with a current ANL-W operator; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 19, 2015; SRDB Ref ID: 

147014 

• Personal Communication, 2015g, Personal Communication with a former ANL-W supervisor; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 20, 2015; SRDB Ref ID: 

147015 

• Personal Communication, 2015h, Personal Communication with a former ANL-W manager; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 20, 2015; SRDB Ref ID: 

147016 

• Personal Communication, 2015i, Personal Communication with a former ANL-W manager; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 20, 2015; SRDB Ref ID: 

147017 

• Personal Communication, 2015j, Personal Communication with a former ANL-W information 

staffer; In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 20, 2015; SRDB 

Ref ID: 147019 

• Personal Communication, 2015k, Personal Communication with a former ANL-W manager; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 20, 2015; SRDB Ref ID: 

147020 
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• Personal Communication, 2015L, Personal Communication with a former ANL-W trainer; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 21, 2015; SRDB Ref ID: 

147011 

• Personal Communication, 2015m, Personal Communication with a current ANL-W operator; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 21, 2015; SRDB Ref ID: 

147012 

• Personal Communication, 2015n, Personal Communication with a former ANL-W engineer; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 21, 2015; SRDB Ref ID: 

147013 

• Personal Communication, 2015o, Personal Communication with a former ANL-W pipefitter; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 21, 2015; SRDB Ref ID: 

147022 

• Personal Communication, 2015p, Personal Communication with a former ANL-W supervisor; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; July 22, 2015; SRDB Ref ID: 

147028 

• Personal Communication, 2015q, Personal Communication with a former ANL-W manager; 

In-person Interview by NIOSH and ABRWH; July 22, 2015; SRDB Ref ID: 147249 

• Personal Communication, 2015r, Personal Communication with a former ANL-W technician 

In-person Interview by [ORAU Team or NIOSH]; July 22, 2015; SRDB Ref ID: 147446 

• Personal Communication, 2015s, Personal Communication with a former ANL-W designer; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; July 22, 2015; SRDB Ref ID: 

147445 

• Personal Communication, 2015t, Personal Communication with a former ANL-W reactor 

operator; In-person Interview by ORAU Team and NIOSH; November 4, 2015; SRDB Ref ID: 

150163 
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4.4 Previous Dose Reconstructions

NIOSH reviewed its NIOSH DCAS Claims Tracking System (referred to as NOCTS) to locate

EEOICPA-related dose reconstructions that might provide information relevant to the petition 

evaluation. Table 4-2 summarizes the results of this review. (NOCTS data available as of January 15, 

2016) 

Table 4-2: No. of ANL-W Claims Submitted Under the Dose Reconstruction Rule

Description Totals

Total number of claims submitted for dose reconstruction 346

Total number of claims submitted for energy employees who worked during the period under 

evaluation (April 10, 1951 through December 31, 1979). 273

Number of dose reconstructions completed for energy employees who worked during the period

under evaluation (i.e., the number of such claims completed by NIOSH and submitted to the 

Department of Labor for final approval). 237

Number of claims for which internal dosimetry records were obtained for the identified years in 

the evaluated class definition 161

Number of claims for which external dosimetry records were obtained for the identified years in 

the evaluated class definition 209

NIOSH reviewed each claim to determine whether internal and/or external personal monitoring 

records could be obtained for the employee.

4.5 NIOSH Site Research Database

NIOSH also examined its Site Research Database (SRDB) to locate documents supporting the 

assessment of the evaluated class. Three thousand nine hundred eighteen (3,918) documents in this 

database were identified as pertaining to ANL-W. These documents were evaluated for their 

relevance to this petition. The documents include historical background on data types, monitoring 

programs, smear sampling, air monitoring, bioassay data, radiological control programs, medical 

monitoring, process materials, and process descriptions. 

4.6 Documentation and/or Affidavits Provided by Petitioners

In qualifying and evaluating the petition, NIOSH reviewed the following documents submitted by the 

petitioners:

• Special Exposure Cohort Petition – Form B, submitted by petitioner representative Atomic 

Weapons Employee Consultants, LLC; submitted November 25, 2014; DSA Ref ID: 120480
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5.0 Radiological Operations Relevant to the Class Evaluated by 

NIOSH

The following subsections summarize the radiological operations at Argonne National LaboratoryWest from January 1949 through December 1979, and the information available to NIOSH to 

characterize particular processes and radioactive source materials. From available sources NIOSH has 

gathered process and source descriptions, information regarding the identity and quantities of each 

radionuclide of concern, and information describing processes through which radiation exposures may 

have occurred and the physical environment in which they may have occurred. The information 

included within this evaluation report is intended only to be a summary of the available information. 

5.1 Argonne National Laboratory-West Plant and Process Descriptions

Since its beginning in 1949 until February 2005, the Argonne National Laboratory-West (ANL-W) 

was operated by the University of Chicago under the AEC\ERDA\DOE‘s Chicago Operations Office. 

In February 2005, the Argonne National Laboratory-West facilities were merged with the rest of the 

Idaho National Laboratory under the DOE’s Idaho Operations Office; the facilities that were still 

operational were collectively named the Materials and Fuels Complex (MFC). 

NOTE: ANL-W was originally known as the “Idaho Division” (ID) of the Argonne National 

Laboratory (ANL) and was also referred to as the “Idaho Site” in ANL documents. For the 

purposes of this evaluation report, the site will be referred to as Argonne National 

Laboratory-West or ANL-W (with the exception of source document titles). 

The ANL-W facilities are physically located on portions of the 890-square-mile Idaho National 

Laboratory (INL) Site, which is located in the high desert of eastern Idaho, west-northwest of the city 

of Idaho Falls.

NOTE: The INL site was originally known as the National Reactor Testing Station (NRTS). The 

AEC established NRTS in 1949 on the site of a 1940s U.S. Navy bombing and artillery range. 

Later, the site was called the Idaho National Engineering and Environmental Laboratory 

(INEEL). On February 1, 2005, INEEL and ANL-W merged to become INL.

The ANL-W facilities primarily occupy two main areas on the INL Site: 

1. EBR-I Complex (a.k.a. ANL West, West Area) (see Section 5.1.1) 

2. EBR-II Complex (a.k.a. ANL West, East Area) (see Section 5.1.2) 

NOTE: The structure of this evaluation report will reflect the above geographical division. 

Figure 5-1 depicts the INL location in the State of Idaho, and the location of ANL-W facilities on the 

INL site (i.e., the EBR-I Complex and the EBR-II Complex). 
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Source: Sehlke, 2010

Figure 5-1: Locations of ANL-W Facilities (EBR-I Complex and EBR-II Complex)
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5.1.1 EBR-I Complex

The EBR-I Complex is located approximately 20 miles west of the EBR-II Complex. The origins of 

all the EBR-I Complex facilities can be directly traced back to Argonne National Laboratory in 

Lemont, Illinois. The facilities were located on the INL reservation due to its remote location. EBR-I 

was commissioned to test the theory of a fuel-breeding reactor. Later, a series of experimental 

boiling-water reactors were built to test steam formation and reactor operating characteristics after a 

steam explosion in Lemont in 1952. The Zero Power Reactor-III facility was part of a large fast 

reactor program initiated at ANL. The Argonne Fast Source Reactor was the last facility built at the 

EBR-I Complex and served as a small-scale reactor for experimental measurements.

EBR-I Complex primary radiological facilities included: 

• Experimental Breeder Reactor-I (EBR-I) (see Section 5.1.1.1) 

• Zero Power Reactor-III (ZPR-III) (see Section 5.1.1.2) 

• Boiling Water Reactor Experiments I-V (BORAX-I – BORAX-V) (see Section 5.1.1.3) 

• Argonne Fast Source Reactor (AFSR) (first location) (see Section 5.1.1.4) 

NOTE: The AFSR was located in the EBR-I Complex until the late-1970s when it was relocated to 

the EBR-II Complex. 

Figure 5-2 shows an aerial photograph of the main part of the EBR-I Complex. 

Figure 5-2: Aerial View of the Main Part of the EBR-I Complex
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Figure 5-3 shows an aerial photograph of the BORAX locations relative to the main part of the EBR-I 

Complex. 

Source: Google Earth

Figure 5-3: Aerial View of the Entire EBR-I Complex
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5.1.1.1 Experimental Breeder Reactor-I (EBR-I)

ATTRIBUTION: Section 5.1.1.1 was completed by Mitch Findley, MJW Corporation. All conclusions 

drawn from the data regarding EBR-I were peer-reviewed by the individuals listed on the cover page. 

The rationale for all conclusions in subsequent sections of this document devoted to EBR-I are 

explained in the associated text.

In 1946, the Manhattan Engineer District (MED) approved a proposal to build the Experimental 

Breeder Reactor No. 1 (EBR-I). Construction began in late 1949 at the National Reactor Testing 

Station in Idaho, now called the Idaho National Laboratory. Early in 1951, a few months before the 

EBR-I facility was completed, nine staff members from the Argonne National Laboratory in Lemont, 

Illinois arrived to install the reactor vessel that was designed and built at Argonne (see Figure 5-4). 

Source: Stacy, 2000

Figure 5-4: Installation of EBR-I Reactor Vessel

The first attempt to operate the new reactor in May 1951 proved to be unsuccessful. It was 

determined that there was not enough fuel in the core to achieve criticality. Acquiring additional 

uranium and refabricating slightly larger fuel rods (in Chicago) took nearly three months. In August 

1951, EBR-I achieved criticality with a core about the size of a football. Four months of low-power 

operation followed (Progress Report, 1953). 
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On December 20, 1951, the first historic EBR-I experiment began. The reactor was started and the 

power gradually increased over several hours. At 1:50 PM, the first usable amount of electricity ever 

generated from nuclear power began flowing from the turbine generator, successfully illuminating

four light bulbs (see Figure 5-5). The next day, the experiment was repeated and sufficient electricity 

was generated to power the entire EBR-I facility (EBR-I, undated). 

Source: Stacy, 2000

Figure 5-5: Four Light Bulbs Illuminated by EBR-1 (Dec. 20, 1951)

EBR-I's real mission was not to show that electricity could be generated by a nuclear reactor, but 

rather, to determine whether scientists' theoretical calculations on fuel-breeding could actually be 

achieved. Early in 1953, analysis showed that EBR-1 was creating one new atom of nuclear fuel for 

each atom it burned. The concept of fuel-breeding had been proven (Summary Report, 1953). Three 

additional cores were developed over the next ten years to increase the breeding ratio. The last core, 

Mark IV, achieved a breeding ratio of 1.27. 

The EBR-I reactor was an unmoderated and heterogeneous design with a thermal output of 

approximately 1 MW at full power. The reactor-tank assembly, and all piping leading from it through 

the shielding, were double-walled (see Figure 5-6). The inner-tank assembly consisted of several 

hexagonally-shaped, stainless-steel subassemblies filled with either fuel or blanket rods. The portion 

of the tank containing the core and inner blanket was only 28 inches long and 15.9 inches ID (interior 

diameter). The vessel weight was supported by a shoulder formed by an increase in diameter at the 

junction of the lower tank assembly. The wall of the lower portion of the inner-reactor tank was 
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5/16-inch thick. An outer Inconel tank with a wall thickness of 1/16-inch served as an emergency 

receiver in the event of an inner-tank failure. The tanks were separated by a system of ribs running 

vertically along the inside surface of the outer tank (Inner Tanks, 1957). 

Source: Radionuclide Concentrations, undated

Figure 5-6: Cut-away View of EBR-I 

The reactor shielding contained a series of horizontal and vertical cross sections. Six horizontal beam 

holes were used to locate various nuclear instruments. A single three-inch hole at centerline elevation, 

and off-center with respect to the core and blanket, passed completely through the shielding. All 

beam holes were steel-lined and closed with concrete-filled steel plugs during operation. An 

extension of the graphite reflector into the concrete-shielded region served effectively as a thermal 

column. Five vertical holes penetrated the graphite reflector from the reactor top. These could be 

used for the irradiation of samples in a thermalized neutron flux. Special thimbles designed for 
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installation in the core or inner-blanket regions were used for the irradiation of samples in a 

fast-neutron atmosphere (Mark IV Terminal Report, 1965). 

Immediately surrounding the lower portion of the reactor tank was the outer blanket (reflector). 

Outside the reflector was an annular layer of graphite, 18 inches thick and 35 inches high. Graphite 

was used in this region to thermalize and to reflect back into the breeding region a fraction of the 

neutrons normally lost through leakage. On the outside of the graphite was a four-inch-thick cast iron 

wall followed by eight feet of concrete (Mark IV Terminal Report, 1965). This thickness of concrete 

reflects additional concrete added in 1952 after it was observed that the dose rates surrounding the 

reactor during operation were higher than expected (Graham, 1959). 

The bottom of the reflector was shielded by four feet of concrete. Radiations penetrating this shield 

streamed into equipment cells located in the basement directly beneath the reactor. Radiations emitted 

upward were absorbed, for the most part, in the coolant and rod extensions. Transmitted radiations 

were additionally absorbed by two shields on the top of the reactor tank. The lower shield consisted 

of two feet of high-density concrete; the upper shield consisted of eight-inch laminate of Masonite and 

iron (Mark IV Terminal Report, 1965). When accessing the reactor, both shields were handled with 

the 20-ton overhead crane (see Figure 5-7).

Source: EBR-I History, 1979, PDF p. 11

Figure 5-7: View of EBR-I 
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To permit the efficient removal of heat under high-power-density conditions (upwards of 180 

kW/liter), and to minimize neutron moderation, a sodium-potassium (NaK) coolant was used. Both 

the primary and secondary systems used NaK which, with a melting point of -12.5°C, was liquid at 

room temperature. Under the effect of gravity, the NaK coolant flowed from an elevated supply tank 

through the inner blanket and core, through a primary-secondary heat exchanger, and into a receiving 

tank in the basement (see Figure 5-8). A pump, operating at a capacity slightly greater than the main 

coolant flow, returned the coolant to the gravity supply tank. An overflow system, connected from the 

supply tank to the receiver, guaranteed a constant level of delivery. From a safety viewpoint, the 

elevated supply tank provided approximately nine minutes of full coolant flow in the event of pump 

failure. If necessary, flow from the gravity tank could be throttled to provide up to eight hours of flow 

at a rate sufficient to remove decay heat (Mark IV Terminal Report, 1965). 

Source: Mark IV Terminal Report, 1965

Figure 5-8: EBR-I Reactor Systems

Heat from the secondary coolant (also NaK) was removed in either of two ways: (1) through the 

generation of steam used to drive a conventional turbine-generator; or (2) through a fan-cooled load 

dissipater. Because the primary coolant was intensely radioactive during operation and immediately 

following shutdown (due to activation of the coolant), all primary components were enclosed in 

concrete-shielded cells. On the other hand, under normal operations, the secondary coolant was not 

radioactive and required no shielding. 
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Of the total thermal power generated in the system, approximately 16% was produced in a massive 

movable reflector that surrounds the primary reactor tank. The reflector consisted of an assembly of 

84 stainless-steel-clad, keystone-shaped bricks of natural uranium weighing 100 lbs. each (Inner 

Tanks, 1957). The entire assembly (weighing approximately five tons) was mounted on a pedestal 

that could be raised or lowered in toto relative to the core. A forced circulation of air through cooling 

holes in the blanket bricks removed heat generated through fission and gamma absorption. There 

were twelve stainless-steel-clad uranium rods that served as a coarse control of reactivity and as a 

means of rapid reduction in reactivity in the event of a scram signal. 

The reactor air supply and outer-blanket-cooling and shield-cooling exhaust systems were 

interconnected. A blower provided the supply air, which was taken from the outside the building and 

forced through an electrostatic filter. Cooling air for the shield was pulled from the supply plenum by 

a centrifugal blower located on the outlet side of the shield-cooling system. Before its release through 

the exhaust stack, the cooling air was passed through a bank of high-efficiency filters. The reflectorcooling system also received air from the same plenum. A positive-displacement blower drew air 

downward through cooling holes, through the reflector-support shielding, through telescopic tubes in 

the elevator structure, and into a plenum chamber. From there, it was drawn through high-efficiency 

filters and released through the stack. 

All NaK storage tanks, both mechanical circulating pumps, and the top of the reactor operated under a 

controlled atmosphere of purified argon at a pressure of approximately 5 lb./in2

. A similar system 

operating at a slightly higher pressure was associated with the secondary NaK system. Feed gas, 

consisting of welding-grade argon, was passed through pressure-reducing valves to a purifier that 

consisted of a stainless-steel-mesh-packed column of heated NaK. Impurities of oxygen and water 

were removed through chemical reaction with the NaK. 

EBR-I history can be traced by its four core loadings. Cores one through three had cores between 50 

and 70 kg of enriched uranium, while the fourth core was plutonium. 

• Mark I: This core was 93% enriched unalloyed uranium in stainless-steel jackets. It was processed 

at the Chemical Processing Plant (CPP) in 1965. 

• Mark II: This core was 98% enriched uranium with a 2% zirconium alloy. On November 29, 

1955, this core experienced a partial meltdown. Interestingly, EBR-I was originally scheduled to 

be shut down by 1956 because researchers thought all useful experiments had been carried out at 

EBR-I; the design of the ZPR-III reactor (operational in 1955) offered greater versatility. As a last 

EBR-I experiment, it was decided to take measurements of the transient temperature coefficients, 

which involved briefly shutting off the NaK coolant. The meltdown occurred because the reactor 

operator used the slower control rods instead of the fast-acting shut-off rods (as intended) to scram 

the reactor. The two-second time difference in using the slower control rods allowed for an 

overshoot in temperature and the resultant partial core meltdown (Radionuclide Concentrations, 

undated). Some of the blanket material melted as well, resulting in an average enrichment of 69% 

for the melted core assembly. Figure 5-9 shows the core assembly after partial removal of the 

lower blanket section. A custom dissolution of the core was performed at the Idaho Chemical 

Processing Plant (ICPP) (IN-1088). 
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Source: Kittel, 1957

Figure 5-9: View of Melted Mark II Core Assembly

• Mark III: This core was also composed of 98% enriched uranium with a 2% zirconium alloy and 

was used to prove that there was nothing intrinsically unsafe about a breeder reactor. 

• Mark IV: This core was composed of 98.6% plutonium, 1.25% aluminum, and some impurities. 

The fuel was fabricated and canned at Rocky Flats in a joint effort with ANL-E (Fuel Elements, 

1956). The purpose of the study was to better understand the properties of a plutonium core and 

determine if the expected breeding improvements over a uranium core could be achieved. Eleven 

of the plutonium fuel slugs were sent to the Metallurgy Division in Chicago for examination. 

In 1966, President Lyndon B. Johnson designated EBR-I as a Registered National Historic Monument 

with the understanding that custody of the facility would be transferred to the National Park Service 

after operations. An EBR-I Complex Decontamination and Decommissioning (D&D) Program was 

prepared in September 1973 to make the EBR-I Complex safe for the public prior to transfer to the 

National Park Service. The primary effort consisted of removing and processing 5500 gallons of NaK 

coolant from the EBR-I reactor system. D&D was performed throughout the EBR-I complex and was 

completed on June 13, 1975 (Kendall, 1975; Brown, 1987). 
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5.1.1.2 Zero Power Reactor-III (ZPR-III)

ATTRIBUTION: Section 5.1.1.2 was completed by Mitch Findley, MJW Corporation. All conclusions 

drawn from the data regarding ZPR-III were peer-reviewed by the individuals listed on the cover 

page. The rationale for all conclusions in subsequent sections of this document devoted to ZPR-III 

are explained in the associated text.

From 1955 to 1990, more than a hundred Zero Power Reactor (ZPR) critical assemblies were 

constructed by Argonne National Laboratory (Lemont, IL) to support fast reactor development, 

validate nuclear data, and establish criticality safety benchmarks. ZPR-III and ZPPR were the 

“chronological bookends” for these critical assemblies and were located at ANL-W. The ZPR-III 

facility began operations in 1955 and ceased operations in 1970. ZPPR began operations in 1969 and 

was the last of the ZPR critical assemblies to be shut down. 

The primary purpose of the ZPR-III facility was to establish the feasibility of a sodium-cooled, 

unmoderated fast-power breeder reactor for the production of electrical power, and to facilitate the 

design and operation of the then-prototype Experimental Breeder Reactor-II (EBR-II). In addition, 

the cores for the Fermi, Rapsodie, and the Southwest Experimental Fast Oxide Reactor (SEFOR) 

reactors were originally mocked up in ZPR-III (Stacy, 2000). There were a total of 63 separate 

assemblies during the operating history of ZPR-III. (An assembly is a critical reactor configuration - 

typically a mock-up of a particular reactor design.) Some assemblies existed for short periods of time; 

they were loaded to critical, a short series of measurements were performed, and the assembly was 

unloaded. Other assemblies lasted much longer; they were entire programs with several major phases

and hundreds of experiments (LeSage, 2001). 

The ZPR-III fast critical facility was a single-story concrete building located in the EBR-I Complex 

(building designations ANL-713 and RTF-601) (Long, 1961). The building consisted of a high-bay 

assembly room where ZPR-III was located, a control room, work room, fuel storage vault, laboratory 

rooms, and offices (see Figure 5-10). The assembly room was 45.5 ft. long by 42 ft. wide by 29.5 ft. 

high. It was kept at a negative air pressure of two inches of water. An airlock separated the assembly 

room from the rest of the facility. The facility walls were composed of two to three feet of concrete 

for shielding purposes. The wall separating the assembly and control rooms was reinforced to five 

feet of concrete to an elevation of 12 feet (Cerutti, 1956). 
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Source: Long, 1961

Figure 5-10: ZPR-III Floor Plan

The critical assembly machine was a horizontal split-table apparatus consisting of a large steel 

platform upon which two tables (or carriages), one stationary and the other movable, were positioned 

(see Figure 5-11). The matrix assembly was composed of stainless steel or aluminum square tubes, 

nominally 2.18 inches square, which were arranged horizontally on each carriage to form a 31-row-by 

31-column square “honeycomb” matrix (roughly 67 inches by 67 inches). The 31 x 31 array of matrix 

tubes was held together by the steel structural members of the carriage. The movable carriage allowed 

for the assembly to be split in half - both as a safety measure and for convenience in loading and 

unloading. A matrix position in each core half was specified by three parameters: matrix half (S for 

stationary or M for mobile), row letter (A-Z and AA-EE starting from the top), and column number 

(1-31 starting from the left looking from the movable half towards the stationary half) (see Figure 

5-12). Except during reactor operation, the tables were separated by five feet. For reactor operation, 

the movable table was driven against the stationary tale by a screw mechanism (Cerutti, undated). 

Free access of authorized personnel to the Assembly Room was strictly prohibited during operations. 
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Source: ZPR-III, 2010, PDF p. 5

Figure 5-11: ZPR-III Critical Assembly Machine
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Source: ZPR-III, 2010, PDF p. 9 

Figure 5-12: Example of a ZPR-III Core Loading

Each half of the assembly contained a multiple-curie polonium-beryllium source unless there was an 

inherent source like Pu-240 in the core. The source was driven in or out of the assembly by a motor 

drive. Each half also initially contained five safety-control rods. On a scram, the rods were driven by 

compressed air (Cerutti, 1956). Later, five additional safety and control rods were added for 

plutonium cores. 

ZPR-III had no system to cool the matrix loading until the mid-1960s. At that time, a rudimentary 

forced-air cooling system was devised because plutonium fuel plates came into use which contained a 

substantial fraction of heat-emitting Pu-240. Before plutonium fuel was used at ZPR-III, there was 

only one thermocouple per assembly half; five more thermocouples per half were added when 

plutonium fuel was employed. 
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Source: Cerutti, Jan1956

Figure 5-13: Typical Loaded ZPR-III Fuel Drawer

The basic building blocks for the core region were composed of 1/8-inch-thick plates two inches high

of various lengths that were loaded into 2-inch by 2-inch by 15-3/16 inch fuel drawers (see Figure 

5-13) (Cerutti, 1956). Fuel-drawer loading took place in the work room adjacent to the vault where 

fuel plates were stored. The loading was performed under the direction of the project supervisor or 

chief physicist (Operations Manual, 1956). The desired average composition was achieved by loading 

the matrix with drawers containing rectangular plates of different materials (e.g., highly-enriched 

uranium, depleted uranium, graphite, plutonium). The uranium plates had a 1-mil-thick Teflon-based 

protective coating (called Kel-F) colored red with an iron oxide powder in the case of the enriched 

uranium. All plutonium plates were clad in stainless steel. The loading of uranium plates was 

performed on a downdraft table whereas plutonium loadings were done in a small glovebox that was 

moved to the table (Personal Communication, 2015c). 

Before each drawer was loaded, a drawer-loading chart was made as a permanent record. Each 

drawer was confirmed for correct loading by a checker before it was removed from the work room. A 

special cart designed to carry no more than two fuel drawers was used to take drawers to the assembly 

room for assembly-loading under the direction of the project supervisor or chief physicist. A master 

assembly loading chart (also known as a matrix loading map), showing a cross-section of each half of 

the assembly, was kept in the control room. In each of the assemblies, a given matrix position had 

two drawers, a front drawer and a back drawer (the front drawers in the stationary and moveable 

halves were adjacent to the matrix interface between halves). Unloading was the reverse of the 

loading procedure with the custodian of the fissile material checking all material back into the vault as 

it was unloaded (Long, 1961).

The first reactor built in ZPR-III was assembled on October 18, 1955 with the assembly going critical 

on October 20. The critical mass was composed of 142.4 kg of U-235 - a core composition roughly 

that of the then-proposed EBR-II (Cerutti, 1956). Critical masses of the U-235 fuel assemblies ranged 

from 27 to 575 kg during ZPR-III’s operating life (Long, 1963). Although ZPR-III used uranium fuel 
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plates exclusively during its earliest years, the facility was built with the anticipation of plutonium 

loadings so that few facility modifications would need to be made (Long, 1963). On June 28, 1961, 

the first sub-critical Pu-239 loading in ZPR-III began (Safety Monthly Report, 1961). Plutonium 

plates used at ZPR-III were composed of 95% Pu-239, 4.5 % Pu-240, and 0.5% Pu-241 (Long, 1963; 

Personal Communication, 2015c). In addition, MOX fuel was used for some assemblies, such as the 

SEFOR (Southwest Experimental Fast Oxide Reactor) assemblies. These assemblies were used to 

conduct tests for the experimental sodium-cooled test reactor located in northwest Arkansas and 

operated from 1969 to 1972 (when the program ended) (Plutonium Loading, 1965). In April 1969, 

ZPR-III began shipping fuel plates to ZPPR as it began winding down operations (Personnel Surveys, 

1970). The reactor was put in cold standby in November 1970 and was later moved into the EBR-I 

facility for display. 

5.1.1.3 Boiling Water Reactor Experiments I-V (BORAX-I – BORAX-V)

ATTRIBUTION: Section 5.1.1.3 was completed by Mike Mahathy, Oak Ridge Associated Universities. 

All conclusions drawn from the data regarding the BORAX Reactors were peer-reviewed by the 

individuals listed on the cover page. The rationale for all conclusions in subsequent sections of this 

document devoted to the BORAX Reactors are explained in the associated text.

At the Argonne National Laboratory (in Lemont, Illinois) on June 2, 1952, a critical assembly mockup 

of a reactor was accidentally placed into a prompt critical reaction resulting in a steam explosion 

within the assembly. The critical assembly was damaged. Four operators received considerable 

radiation exposure. Nonetheless, an ANL staff member suggested that direct boiling reactors might be 

practical in that steam formation would help stabilize the reaction. On July 23, 1952, ANL authorized 

the design and construction of the first boiling water reactor to test that theory and reactor operating 

characteristics. 

Five reactor configurations were built at the INL and were known as the BORAX (Boiling Water 

Reactor Experiment) series (NARA 071; Dietrich, 1954). The BORAX reactors were part of the 

EBR-I Complex. A control building for BORAX-V was constructed about one-half mile from that 

reactor (Borax-V Description, 1961). Figures 5-1 and 5-3 show the location of the BORAX reactors 

on the INL site and relative to the main part of the EBR-I Complex. The five BORAX reactors were 

built and operated in the periods listed in Table 5-1. Detailed results of the experiments are described 

in the following subsections. 

Table 5-1: BORAX Reactors Operational Periods

Reactor Dates of Operation Source

BORAX-I July 1953 to July 1954

(summer and fall months only)

Borax Experiments, 1954

BORAX-II October 1954 to March 1955 Reactor Quarterly, 1954, SRDB 143253

BORAX-III July 1955 to April 1956 Reactor Quarterly, 1956a

BORAX-IV December 1956 to June 1958 Reactor Quarterly, 1956b, Defective Fuel 

Experiment, 1959

BORAX-V March 1962 to August 1964 Progress Report, Sep1964
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BORAX-I 

In 1952, ANL designed an experimental reactor to be used for determining self-limiting characteristics 

of water-cooled reactors and the operating characteristics of boiling reactors. The reactor was 

fabricated at ANL and was shipped to and set up at ANL-W in early summer 1953. The reactor could 

only be used in summer and warm fall months because it was not totally enclosed within a structure.

The first BORAX reactor was 2730 feet northwest of the entrance to EBR-I. The reactor was 

controlled from a trailer placed just outside the EBR-I entrance; during BORAX-I operation, an 

exclusion radius of one-half mile was maintained around the reactor (Dietrich, 1954). Figure 5-14 

shows a drawing of the BORAX-I site plan.

Source: Borax Experiments, 1954

Figure 5-14: BORAX-I Site Plan

The BORAX-I reactor core was built up from a lower grid into which were placed a combination of 

twenty-six fuel assemblies and ten non-fuel (dummy) plugs. Criticality was reached with the 26 fuel 

assemblies; the dummy plugs were used to fill the core grid to permit relatively-constant geometry for 

water flow. Fuel assemblies were made of curved fuel plates containing a U-235-aluminum-alloy 

core that was clad with aluminum. Each assembly contained 18 fuel plates joined to aluminum side 

plates to form units about three inches square. The fueled portion of the fuel plates was two feet in 
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length (Dietrich, 1954). The core was divided into quadrants with control rods located in the 

openings. The grouped rods were raised to increase reactivity but could also be rapidly forced 

downward by springs for reactor shutdown. The central control rod was moved downward to increase 

reactivity but could be ejected downward for rapid reactivity increases as needed by experiment 

criteria.

The reactor and associated mechanisms and structure are shown by vertical diagram in Figure 5-15. 

The reactor was placed in a 20-foot deep shielded tank sitting on a concrete slab ten feet below ground 

level. The reactor core was located approximately three feet below ground level. Sand and gravel 

were piled up around the shield tank at about a 45-degree angle which provided at least eleven feet of 

sand and gravel shielding for workers situated at ground level. That arrangement permitted area 

access at all times immediately after termination of a run. Sometimes when it was necessary to work 

on the reactor, the shield tank was filled with water; sub-tolerance radiation levels were reached with a 

few hours decay. Control rods were moved and positioned by an external drive mechanism. 

Source: Borax Experiments, 1954

Figure 5-15: BORAX-I, Vertical Diagram
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Through the summer and early fall of 1953, BORAX-I was put through a series of intentional 

excursions of periods shorter than 25 milliseconds, as well as periods of steady boiling operation; 

these latter tests covered operating pressures up to 130 psig (pounds per square inch, gauge). These 

excursions caused steam to escape from the reactor, somewhat resembling a geyser (Figure 5-16). 

While it was not possible to determine all the characteristics of boiling in such a short time, 

researchers at ANL-W did gain conclusive proof that steam formation was an effective and rapid 

power-limiting process. They further deduced that steam would protect properly-designed reactors 

against short millisecond reactivity excursions (Transients, 1954).

Source: Till and Chang, 2011

Figure 5-16: Steam from Excursion of BORAX-I 
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By the summer of 1954, ANL-W staff had designed a replacement for BORAX-I. However, they 

proposed that before the first reactor was replaced, it be subjected to a single, very quick, destructive 

excursion to determine the inherent safety under extreme conditions. After discussion with the AEC 

Reactor Development Division and the Advisory Committee on Reactor Safeguards, ANL-W staff 

decided that such a final experiment was worthwhile. Consequently, BORAX-I was deliberately 

destroyed in July 1954. ANL-W staff encountered a hitch at the start of the destructive test. A test 

run was successful without complication. However, upon inspection, they learned that a fuel plate had 

bowed out so that it was rubbing against the control rod. The staff was under pressure to start the final 

test. Looking down into the reactor, they could see that the fuel subassemblies were giving off a 

Cerenkov glow. Even so, staff members lifted the bowed fuel subassembly up out of the reactor, 

fixed it, and put it back in place. They vacated the reactor and were successful in starting the test. 

At the moment of explosion, fuel plate fragments were scattered for a distance of 200-300 feet, but 

there was no widespread dangerous dispersal of radioactive materials (ANL-W History, 2006). Gold 

foils had been placed at various locations in the reactor core so that staff could determine the power 

the reactor attained before exploding. These foils were about one centimeter square and attached to 

the outside of the fuel subassemblies. The foils were embossed with numbers to record their 

placement location in the reactor. Over several days, reactor workers were rotated into the reactor pit 

to recover the gold foils. Site staff also retrieved materials and usable equipment from the reactor and 

surrounding area over the next two months (Haroldsen, 2008). 

The final test revealed that the predictions of total energy and fuel plate temperatures were 

considerably too low. Instead of melting a few fuel plates, the test melted a major fraction of the 

entire core. The discrepancy was attributed to the uncertainties of extrapolation. The results of this 

energy liberation in regard to peak pressures and explosive violence were in a region without previous 

experimental data.

ANL published a video in March 2013 about BORAX-I called Borax - Safety Experiment on a 

Boiling Water Reactor. This includes actual footage taken during experiments and provides more 

detail on reactor installation and operation (ANL, 2013). 

BORAX-II

BORAX-II was designed and built by ANL and operated at ANL-W to replace BORAX-I. The goal 

was to extend the BORAX-I experiments on the transient and steady-state operation of boiling 

reactors with a new reactor that more closely approximated the characteristics of a practical power 

reactor operating on the boiling principle. BORAX-II was larger than BORAX-I and designed for 

operation at higher pressure (300 psig). It was installed in August 1954 on a new site about 200 yards 

east of the BORAX-I site (see Figure 5-3). The first BORAX-II operation was conducted in October 

1954 (Reactor Quarterly, 1954). Figure 5-17 shows a photo of BORAX-II.

With BORAX-II, ANL tested new core combinations using varying enrichments of U-235 in the 

metal fuel plates. The boiling water system was operated at 300 psig, making it essentially a power 

experiment. Initially, BORAX-II had no turbine generator; thus, no electricity was produced. As with 

BORAX-I, the energy produced was released in the form of steam. 

The reactor tank was made of stainless steel and was installed completely below ground level in a 

concrete-lined pit that could be flooded with water for shielding for reactor work conducted during 
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shutdown. A pit was provided with a removable lid of three-foot-thick concrete. The control rod 

mechanism was mounted above the pit with extension rods passing through the concrete to drive the 

control rods. Much of the ancillary equipment used outside the reactor tank was originally used on 

BORAX-I. Those items were decontaminated and repaired after the final BORAX-I destructive 

experiment. Several new instruments were required.

A machinery pit containing feed pumps, water preheat circuitry, and steam controls was located 

adjacent to the reactor pit but shielded from the pit by five feet of concrete. The deionizer system for 

the reactor water, and the nuclear and electronic instrumentation for recording results of transient 

experiments were mounted at ground level. The entire installation was covered by a 40-foot by 

60-foot Butler building. The reactor was operated remotely from a control trailer one-half mile south 

of the reactor. 

Source: Till and Chang, 2011

Figure 5-17: BORAX II

The reactor core was mounted near the bottom of the reactor tank. The core consisted of an adjustable 

number of modified MTR-type fuel elements held in a closely-packed assembly by an aluminum grid. 

The core was completely submerged in water that served as coolant, moderator, and reflector. Like 

BORAX-I, water was circulated in the reactor tank during reactor operation by natural convection 

(Reactor Quarterly, 1954). The active sections of the fuel elements had the exact external dimensions 

as standard MTR fuel elements, but they contained only ten fuel plates, with a total loading of 93 g U235 per element. The reactor core was divided into quadrants by four pairs of cadmium control rods 
SEC-00224 02-18-16 ANL-W 

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with a cadmium control rod at the center of the core. The five control elements were operated 

independently. 

BORAX-III

In 1955, ANL replaced the reactor core and added a turbine-generator to BORAX-II to test power 

distribution and whether turbine contamination would be a significant problem in a boiling water 

reactor. The revised reactor configuration was designated BORAX-III. The core consisted of 

uranium fuel enriched to about ninety percent U-235. The uranium was alloyed with 

aluminum. The uranium-aluminum alloy was then clad with aluminum. Plate-type fuel

elements were used. The fuel plates had a short span to avoid distortion from pressure 

resulting from steam formation inthe coolant channels during transient tests (Revised Reactor 

Configuration, 1956; Fuel Plate Fabrication, 1958). The reactor vessel, made of stainless steel, 

was unchanged from BORAX-II. New quadrant control rods fabricated from boron-hafnium were 

added to replace the BORAX-II control rods. The center control rod made from cadmium was not 

replaced. The reactor, control rod mechanisms, water control systems, canal for spent fuel storage, 

and steam piping was housed in the sheet metal building constructed for BORAX-II (Control Rods, 

1957).

The mission of BORAX-III was to generate electrical power for Arco, Idaho on a one-time basis. A 

turbine generator, condenser, and equipment to return cooled water to the reactor were installed in a 

new building. The steam pipe in the reactor building outlet was connected to the turbine via a 

switching mechanism (see Figure 5-18). Water was condensed and pumped through an adjacent 

cooling tower, then back to the reactor building for use as feed water (Water Pumps, 1956). Figure 

5-19 shows a diagram and photo of the BORAX-III facility. The steam could also be diverted as 

outdoor exhaust by closing the turbine switch. 

Source: ANL, undated

Figure 5-18: BORAX-III Turbine
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Source: Till and Chang, 2011

Figure 5-19: BORAX-III Facility

ANL-W workers were successful in preparing and operating BORAX-III to generate electricity for an 

entire city for the first time anywhere in the world. On July 17, 1955, electricity produced by 

BORAX-III supplied the town of Arco, Idaho with its entire supply of electrical power (500 kW), 

generated 500 kW to power the BORAX facility, and 1000 kW to power the Central Facilities Area at 

INL. 

ANL West continued to operate the BORAX-III reactor until April 7, 1956 to test operating 

characteristics during steady-state operation and to identify fission and activation product 

contaminants in the plant’s air and water (Reactor Steady-State Operation, 1956). The core was 

removed in June 1956 in preparation for fuel loading for BORAX-IV (Reactor Quarterly, 1956a). The 

BORAX-III reactor was operated for a total of 1170 hours. 
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BORAX-IV

ANL-W replaced the BORAX-III core with a core of a maximum seventy-two ceramic uraniumthorium oxide fuel elements to test and demonstrate the feasibility of stable operation with a fuel that: 

(1) could operate at higher temperatures than the uranium core; and (2) was considered less reactive 

with water coolant in case of cladding rupture (Fuel Fabrication, 1957; Fuel Element Testing, 1958; 

Reactor Quarterly, 1956b; Fuel Element Testing, 1958). The new reactor configuration was known as 

BORAX-IV. Other than the core, all other BORAX-III components were unchanged with BORAXIV. Like the BORAX-III reactor, BORAX-IV was operated at 300 psig and had the potential to 

produce 2.5 MW of electricity. Operation of BORAX-IV began on December 3, 1956. 

BORAX-IV was first brought to criticality under atmospheric pressure, with water temperatures 

ranging from cold to boiling (93 degrees C at the elevation of BORAX-IV). It was operated with this 

core until April 17, 1957. A revised core with uranium-thorium oxide elements was installed in late 

April 1957. BORAX-IV was operated with that core at 300 psig and 216 degrees C from May 1957 

intermittently through December 5, 1957. BORAX-IV also produced measurable quantities of the 

artificial, thorium-derived fuel, U-233. 

The core was again revised in order to increase the maximum power for continuation of stability 

studies. Boron steel poison rods were removed with ten new elements added to the periphery of the 

core. On February 19, 1958, the reactor was restarted with the revised core. BORAX-IV was 

operated during early 1958 to evaluate the effect of operating a direct-cycle boiling water reactor with 

a fuel element defect, and for locating defective elements in the core. The initial test performed in 

February was done with simulated ruptured elements (Request for Pellets, 1957; Defective Fuel 

Experiment, 1959). On March 11-12, 1958, the BORAX-IV reactor was operated at 2.4 MW, even 

though a large number of fuel elements contained defects through the cladding (Borax-IV Test, 1959). 

Measurements made during the reactor operations included radioactivity levels of the steam plant 

equipment; quantitative determination of the fission gases Xe-138 and Kr-88 that were released 

through the air ejector; analyses of reactor water, condensed steam before the turbine, and condensed 

steam for fission products; and area contamination downwind from the reactor. BORAX-IV was 

operated until June 1958. Clean-up and salvage operations began in July and continued through 1959 

(Safety Monthly Report, 1958; Air Sample Data Sheets, 1965). Some equipment was salvaged for use 

with BORAX-V. The reactor vessel was entombed in place using sand capped with concrete (Cleanup Request, 1959; Arave, 1992). Most other equipment and facilities were removed and buried in 

1960 and 1961. Fuel plates were shipped to CPP at INL for processing. Personnel involved in the test 

were monitored for external radiation, and for intakes of fission products by urinalysis; urinalysis 

detected no activity above background (Graham, 1959).

BORAX-V 

In 1957, the concept for BORAX-V was derived by ANL’s Reactor Engineering Division to test 

nuclear superheating concepts and to advance the art of boiling-water-reactor design. The greatest 

departure in design for BORAX-V was in the superheater assembly.

Some of the plant equipment used in the previous BORAX experiments was incorporated in the 

BORAX V facility, but a new reactor building and reactor assembly with new reactor vessel were 

constructed. A reactor control building was constructed about one-half mile away (Borax-V 
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Description, 1961). Figure 5-20 shows a drawing of the reactor facility; the control building shown in 

the lower right is one-half mile away from reactor (Borax-V Design and Hazards, 1961).

The reactor vessel was a cylinder with ellipsoidal heads made of carbon steel clad internally with 

stainless steel. The top portion of the vessel was used as a steam dome. The core is centered about 

four feet from the vessel bottom. Control rods made of Boral and canned in stainless steel were 

manipulated from below the vessel (Uranium Core Configurations, 1961; Holtz, 1964).

Source: Brunson, 1959

Figure 5-20: BORAX-V Facility

BORAX V was designed to support three separate uranium core configurations: (1) a boiling core 

without superheater; (2) a boiling core with a centrally-located superheater; and (3) a boiling core with 

a peripherally-located superheater. In each case, ANL assumed it was possible to operate the reactor 

with either natural or forced circulation of water through the core. Water served as the moderator in 

both the boiling and superheater regions of the core and as coolant in the boiling region. The 

superheater was cooled by steam (Uranium Core Configurations, 1961; Holtz, 1964).

Each of the three cores was 24 in. high with a maximum diameter of 39 in. (depending on the number 

of fuel assemblies loaded). The central superheater contained twelve fuel assemblies in the central 

position, while the peripheral superheater contained sixteen fuel assemblies, and four in the middle of 

each outer row (Borax-V Design and Hazards, 1961). The three core structures were readily 

exchangeable after removal of fuel and control rods. A core structure was easily withdrawn from the 

vessel as a unit by a remote- controlled crane and stored in the dry-storage pit.
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Each of the three cores was used in the reactor. Subcritical experiments were performed on each core. 

Two configurations of boiling core without a superheater operated intermittently from March 1962 

through February 1963. Two configurations of boiling core with a centrally-located superheater 

operated between April and November 1963. The boiling core with a peripherally-located superheater 

was operated between April and July 1964. At the end of July 1964, some fuel elements were 

replaced with prepared defective fuel elements to test power generation using defective fuel. The core 

was operated until August 14, 1964, producing 266 MW with the defective fuel. That final 

experiment in BORAX V demonstrated that negligible contamination to turbo-generator equipment 

resulted from operating with an experimentally-defective fuel element in the superheated core. The 

reactor was shut down and disassembled by the end of September 1964. The vessel was stored in the 

basement of the reactor building. The vessel used in BORAX-II/III/IV was also transferred to that 

basement for long-term storage (Progress Report, Sep1964).

5.1.1.4 Argonne Fast Source Reactor (AFSR) (1st location: 1959-1970) 

ATTRIBUTION: Section 5.1.1.4 was completed by Mike Mahathy, Oak Ridge Associated Universities. 

All conclusions drawn from the data regarding AFSR (first location) were peer-reviewed by the 

individuals listed on the cover page. The rationale for all conclusions in subsequent sections of this 

document devoted to the AFSR (first location) are explained in the associated text.

The AFSR was designed and built in 1958 near EBR-I and ZPR-III. AFSR was placed in operation in 

October 1959 with a design power of one kilowatt. AFSR was a laboratory-scale reactor designed to 

supplement ANL-W’s existing facilities by providing a readily-available source of reproducible fluxes 

of fast and thermal neutrons for developing, testing, calibrating, and standardizing various counters; 

preparing radioactive metallic foils for developing counting and radiochemical techniques; testing 

complex experimental systems before operation in other reactors; and developing potential 

experiments in the fast reactor field (Brunson, 1959; IOO Information Packet, 1962).

AFSR was housed in a prefabricated Butler-type building measuring 32 x 32 x by 20 feet high. The 

reactor incorporated a cylindrical core of solid, highly-enriched uranium approximately 4.5 inches in 

diameter by 4.25 inches high. The core was divided into upper and lower sections. The upper section 

constituted about three-fifths of the critical material. It consisted of two uranium discs separately 

canned and held in position by the upper cage. There was a horizontal one-half-inch diameter "glory 

hole" at the approximate reactor center line. The lower section of the core and the supporting blanket 

section to which it was attached comprised the safety plug which could be raised and lowered using an 

air cylinder. The lower core section was composed of discs of various thicknesses so that the core 

loading could be varied. Discs used in the lower section were attached to the supporting blanket 

section by means of a second stainless steel cage (Brunson, 1959). 

The reflector was made of solid depleted uranium with a minimum thickness of eight inches. The 

outer form was cylindrical. The reactor was freestanding on the reactor floor inside a shield of highdensity concrete. A graphite thermal column 4 x 4 x 6 feet was provided for experiments. All control 

and safety mechanisms were located in a pit beneath the reactor (Brunson, 1959). A 15-curie 

polonium-beryllium neutron source was located in the reflector. The source was manually retracted 

into the shield by means of a mechanical drive (Brunson, 1959; IOO Information Packet, 1962; 

Summary Report, 1960). Figure 5-21 shows the reactor building layout. Figure 5-22 shows a photo 

of the building. 
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Source: Brunson, 1959

Figure 5-21: AFSR Building Layout (1st location)

There were three experimental holes in the reflector. All experimental and instrument holes through 

the shield were provided with stepped plugs. The reactor biological shield was composed of 

magnetite-aggregate concrete with a density of 215 pounds per cubic foot. The nominal shielding 

thickness was 4.5 feet around and above the reactor. The thermal column was shielded by a one-inch 

steel plate faced with 1/8-inch of Boral. As the sub-reactor pit was not to be occupied during 

operation, the region below the reactor was only partially shielded. A minimum of two feet of 

magnetite concrete was provided between the pit and the floor outside the shield to absorb neutrons 

scattered off the pit floor. 
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Source: Till and Chang, 2011

Figure 5-22: AFSR Building (1st location)

AFSR was shut down in October 1961 when staff discovered deterioration in the lower fuel section. 

After disassembly of the lower section, small pimples were visible under the thin nickel can. The fuel 

was shipped to ANL (Chicago) where replacement cans were fabricated and installed around the core 

pieces. The fuel was returned in two shipments, one by the end of May 1962 and the other in July 

1962 when full use of AFSR was restored (Reactor Operation, 1962). 

Various documents discuss some of the irradiations and tests performed at AFSR (Summary Report, 

1960; Progress Report, 1962; Borax Superheater, 1965; Progress Report, 1965; AFSR, 1965; Progress 

Report, 1967; Annual Report, 1968). To accommodate the close of the EBR-I and BORAX reactor 

areas, AFSR was moved to a new location adjacent to the ZPPR facility in the EBR-II Complex in fall 

1970. 

A small, consistent number of people (4 to 8) periodically worked at AFSR. The work area for 

personnel is identified in each individual’s monitoring records (AFSR Quarterly Reports, 1971).
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5.1.2 EBR-II Complex

EBR-II Area primary radiological facilities during the period under evaluation (April 10, 1951 to 

December 31, 1979) included: 

• Transient Reactor Experiment and Test Facility (TREAT) (see Section 5.1.2.1) 

• Experimental Breeder Reactor-II (EBR-II) (see Section 5.1.2.2) 

• Fuel Cycle Facility (FCF) (see Section 5.1.2.3) 

• Fuel Assembly & Storage Building (FASB) (see Section 5.1.2.4) 

• Inspection & Test Facility (ITF) (see Section 5.1.2.5) 

• Hot Fuel Examination Facility- North (HFEF-N) (see Section 5.1.2.6) 

• Zero Power Plutonium Reactor (ZPPR) (see Section 5.1.2.7) 

• Laboratory & Operations Building (L&O Building) (see Section 5.1.2.8) 

• Argonne Fast Source Reactor (AFSR) (2nd location, 1970 to 1993 (see Section 5.1.2.9) 

• Miscellaneous Facilities (see Section 5.1.2.10) 

NOTE:The AFSR was located in the EBR-I Complex until the late-1970s when it was relocated to the 

EBR-II Complex. 

Facilities in the EBR-II Complex were not constructed until several years after the EBR-I Complex 

facilities began operation. The first radiological operations in the EBR-II Complex took place when 

the TREAT Reactor went into operation in 1959. TREAT is approximately one mile (1.6 km) outside 

the main part of the EBR-II Complex (Site Plan, 1981). The first occupancy of facilities within the 

main part of the EBR-II Complex occurred in 1960, when ANL received “beneficial occupancy” of 

areas in the EBR-II Power Plant and EBR-II Reactor while the construction contractor finished 

constructing the initial EBR-II Complex facilities. By November 1960, ANL had complete 

occupancy of the EBR-II Power Plant and EBR-II Reactor. On July 10, 1961, the construction 

contractor finished all deficiency items and ANL accepted the completed buildings (Summary Report, 

1961, PDF p. 103). 
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Figure 5-23 shows an aerial photograph of the EBR-II Complex in the early 1960s. 

Source: ANL-PIO, undated

Figure 5-23: Aerial View of the EBR-II Complex (circa early-1960s) 

Figure 5-24 below shows the physical location of each of the EBR-II Complex buildings in 1981. 

NOTE: the distance to the TREAT buildings (upper left corner) is not to scale; they are actually about 

one mile away.
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Source: Site Plan, 1981

Figure 5-24: Map of EBR-II Complex in 1981
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5.1.2.1 Transient Reactor Experiment and Test Facility (TREAT)

ATTRIBUTION: Section 5.1.2.1 was completed by Jason Davis, Oak Ridge Associated 

Universities. All conclusions drawn from the data regarding TREAT were peer-reviewed by the 

individuals listed on the cover page. The rationale for all conclusions in subsequent sections of this 

document devoted to TREAT are explained in the associated text. 

The Transient Reactor Test (TREAT) Facility was an air-cooled, thermal, heterogeneous system 

designed to evaluate reactor fuels and structural materials under conditions simulating various types of 

nuclear excursions. TREAT was designed for two modes of operation: steady-state and transient. In 

steady-state mode, it served as a large neutron-radiography facility (TREAT Manual, 1971). In the 

transient mode, TREAT’s primary objective was to provide quantitative data and visual information 

on the mechanism of the melting of fast reactor fuel elements by nuclear heating analogous to a power 

excursion in a fast reactor core.

The TREAT complex was comprised of a reactor building and a control building located about 4250 

ft. and 1750 ft., respectively, northwest of the EBR-II containment vessel. The site was selected 

because the topography permits an unobstructed view between the two buildings. Construction was 

started in February 1958 and completed in early November 1958. The reactor first achieved criticality 

on February 23, 1959. 

The Reactor Building, shown in Figure 5-25, was an aluminum-sided steel-frame structure featuring a 

high-bay section and adjacent service wing. The high-bay section (35-ft. ceiling) contained the 

reactor, fuel storage pit, instrument room, and basement sub-reactor and equipment rooms. The 

control rod drive mechanisms were located in the sub-reactor room, which measured 16 ft. square by 

l3.5 ft. high. An adjacent equipment room housed auxiliary apparatus (e.g., pumps, storage tanks, 

piping) required for experiments. Access to both rooms was by a stairway from the main floor into 

the equipment room, and through a door into the adjacent sub-reactor room. In addition, there was an 

escape ladder leading from the sub-reactor room to the main floor. 

Fuel transfer was accomplished via a lead-shot and steel-shielded cylinder, known as a coffin. The 

motor and gearing for performing refueling operations were mounted at the top of the coffin and 

controlled from a single portable control box. The base of the coffin was positioned above the desired 

opening in the radial slot of the rotating plug and locked in place with an indexed carriage. The 

installation of control-rod fuel assemblies was accomplished by removing the adjacent fuel assemblies 

and rotating an internal tube within the coffin. Operations within the core were accomplished with the 

aid of a shielding window block in the slot of the rotating shield plug when radiation levels exceeded 

those permissible for direct viewing through the slot. A periscope was also available for viewing 

operations within the core (TREAT Facility, 1971).

The Control Building was a single-story, concrete-block structure that contained the control panels 

and instrumentation for remote-controlled reactor operation. 
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Source: TREAT Facility, 1971

Figure 5-25: TREAT Reactor Building Plan
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In the steady-state mode of operation, TREAT served as a large neutron radiography facility. The 

neutron radiography facility (Figure 5-26) consisted of a neutron collimator and a massive shield for 

positioning the radiographic subject. The collimator was a double-tapered, boral-lined channel 

extending from the edge of the core to the center of the radiography shielding, which was located 

outside the normal reactor shielding. A massive rotating shutter provided off-on control of the 

neutron beam. A timer-controlled air solenoid opened and closed the shutter as required for particular 

samples. The beam covered an area 4 in. x 24 in. at a density of approximately 107 neutrons/cm2

-sec 

when the reactor was operating at 50-90 kW of steady-state power. The cast-lead radiography 

shielding permitted immediate access to the facility while the reactor was operating at steady state and 

a highly-radioactive subject was in position. Subjects to be radiographed were lowered into the shield 

through an opening in the top of the shield. Highly-radioactive subjects were lowered from a coffin 

that was positioned on top of the shield (TREAT Manual, 1971). 

The Mark-I integral sodium loop was intended for experimental use in TREAT meltdown experiments 

of relatively low-melting-point metallic fast reactor fuels contained in flowing sodium. These loop 

experiments were sponsored by the Reactor Analysis and Safety Division (RAS) of ANL-East, and 

were designed to simulate either loss-of-coolant flow accidents or transient-overpower accidents in the 

Fast Flux Test Facility (Cook, 1975). The loops were used successfully and routinely for such tests. 

However, the design pressure of 30 atm at 500°C and steady-state temperature limitation below 400°C 

were not adequate for Liquid Metal Fast Breeder Reactor (LMFBR) oxide-fuel transient safety 

experiments. The counterflow design of the Mark- I loop was required to reduce the loop size to fit 

within a single TREAT fuel-element can. However, this counterflow feature made measurement of 

pressure, flow, and temperature difficult. 

The Mark-II integral loop was used for testing LMFBR-type oxide-fuel elements in the TREAT

Facility under transient conditions of coolant flow, coolant temperature, and fuel element power 

representative of those that may occur in a reactor. The loop was positioned within the TREAT 

reactor so that the test section penetrated into the reactor core region. The apparatus arrangement was 

such that no portion of the loop became a permanent or integral part of the reactor itself, with the 

expectation that the Mark-II loop would not affect reactor operation or reactor control once routine 

provisions were made to compensate for the removal of two reactor fuel elements (Stephens, 1976).

During 1973 and 1974, several Mark-IIA loop experiments were sectioned in the HFEF-South argon 

cell as part of the HFEF post-test support of the RAS-TREAT loop experiments. The purpose of 

sectioning the loops was to reduce the loop experiments into smaller segments for shipment to the 

Materials Science Division (MSD) at ANL-East, where the detailed post-test examination of the loop 

experiments was performed. Following irradiation in TREAT, the loop experiments were transferred 

to HFEF in the HFEF loop cask and transferred into the HFEF Air Cell and finally into the Argon 

Cell. In the Argon Cell, the loop insulation, heaters, neutron filters, and miscellaneous loop outfitting 

were stripped from the loops (Cook, 1975). Further discussion of these processes is reserved for the 

HFEF section of this report. 
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Source: TREAT Manual, 1971

Figure 5-26: TREAT Reactor Plan View

The TREAT facility operated from 1959 to 1994, undergoing 6,604 reactor start-ups and 2,885 

transient irradiations before being placed on standby. TREAT was used to study fuel meltdowns, 

metal-water reactions, interactions between overheated fuel and coolant, and the transient behavior of 

fuels for high-temperature systems. During the first few years of operation, a large number of fuel-rod 

meltdown experiments (85 during the first two years) were performed on unirradiated and irradiated 

fuel, most of them in dry capsules. These were largely phenomenological experiments in that they 

were used to identify and study the basic physical processes occurring during rapid overheating of the 

fuel-rod samples, rather than simulating specific accident conditions. Although sodium-bonded 

uranium-alloy fuel was the predominant type tested, various fuel and cladding materials were also 

studied. By the mid-1960s, integral experiments in flowing-coolant loops were being conducted. 
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In the late 1960s, the major emphasis shifted to testing of oxide fuel for application in the Fast Flux 

Test Facility (FFTF), and eventually, in later commercial fast reactors. The Mark-III sodium loop was 

designed for use in these tests. The original Mark-III loops could accommodate up to seven fuel pins 

of the FFTF design of 5.84 mm diameter. These loops used a "package" concept, in which the whole 

experiment was assembled outside the reactor, inserted into an appropriate location in the core for 

transient irradiation, and then removed for disassembly and post-test examination. Supporting 

facilities were established at both Argonne sites to assemble test trains (using both fresh and 

pre-irradiated pins), to insert the test trains into the loops, to disassemble the loops after the tests, and 

to perform post-test macroscopic and microscopic examinations (Deitrich, 1998). 

Also in the late 1960s, the fast neutron hodoscope was developed. This instrument enabled real-time 

tracking of the location of fissile material within an experiment by counting collimated fission 

neutrons emitted from the fuel. TREAT experiments done with the fast neutron hodoscope were the 

primary source of experimental information to validate fuel-motion models used in safety analysis 

(Deitrich, 1998).

Through the decade of the 1970s, the primary emphasis in TREAT experiments was support of FFTF 

and the Clinch River Breeder Reactor Plant (CRBRP) safety analysis and licensing. For FFTF, the 

primary emphasis was on simulations of transient overpower (TOP) and unprotected loss of flow 

(LOF) accidents. The primary interest was in determining the time and location of fuel pin failure, 

and the nature of the subsequent fuel motion. The principal finding was that there was no 

strongly-compactive fuel motion that would lead to a reactivity excursion that would exacerbate the 

assumed input ramp. The LOF experiments, conducted at essentially constant power with decaying 

flow, also showed an absence of strongly-compactive fuel motion, implying an energetically-benign 

transition into a core melt (Deitrich, 1998).

Free access of authorized personnel was permitted to the Reactor Building during shutdown, critical 

operation, or steady-state operation. The dose rate at normally-occupied locations in the reactor 

building was less than 7.5 mr/hr during 100 kw operation (TREAT Facility, 1971). One exception to 

the free-access policy was that access to the subpile room at any time required a check of radiation 

level, and access to this area during reactor operation required the approval of the Supervisor in 

Charge. A warning sign, operated from the radiation alarm system and indicating high activity in the 

basement, was posted at the basement stairway. Prior to prolonged work in sub-pile room, the Health 

Physics representative was to make a complete survey and post any restrictions in conspicuous places. 

Access to the Reactor Building during transient operation was not permitted. Prior to transient 

operation, the Supervisor-in-Charge cleared the building of all personnel, personally inspected all 

areas in the building, and announced over the public-address system that the Reactor Building was 

being cleared for a transient operation (TREAT Manual, 1971, PDF p. 82). 
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5.1.2.2 Experimental Breeder Reactor-II (EBR-II) 

ATTRIBUTION: Section 5.1.2.2 was completed by Brian Gleckler, Dade Moeller, Inc. All 

conclusions drawn from the data regarding EBR-II were peer-reviewed by the individuals listed on 

the cover page. The rationale for all conclusions in subsequent sections of this document devoted to 

EBR-II are explained in the associated text.

The EBR-II reactor was a second generation Liquid Metal Fast Breeder Reactor that operated from 

September 30, 1961 to September 30, 1994 (Stacy, 2000). The EBR-II reactor was basically a scaledup version of the first generation LMFBR (i.e., EBR-I); it was approximately 20 times larger than 

EBR-I (Stacy, 2000). Another difference was their liquid-metal coolants. EBR-I used a eutectic alloy 

of sodium-potassium metal (NaK); EBR-II used only sodium metal (Stacy, 2000). 

The EBR-II concept was strongly influenced by the limited availability of highly-enriched fuel at that 

time due to military applications having priority. It appeared that these materials would have limited 

availability in the future and would be costly. Consequently, the EBR-II concept was based on the 

need to use all means available to achieve a high power density in the reactor while minimizing the 

total fuel inventory (EBR-II, undated).

In Figure 5-24, the EBR-II reactor is Building 767. The EBR-II reactor building is also known as

ANL-767, EBR-II-601, and MFC-767. Cross-Sections of the various portions of the reactor building 

are depicted in Figures 5-27 and 5-28. 

NOTE: Figures 5-27 and 5-28 were assembled from components of images from two historical source 

documents (Koch, 1964, PDF p. 109; Survey Reports, Aug-Dec1969, PDF p. 62). 

NOTE: The legend included in Figure 5-27 applies to Figure 5-28 as well.
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Source: Koch, 1964

Figure 5-27: EBR-II Reactor Building (side view)
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Source: Koch, 1964, PDF p. 109; Survey Reports, Aug-Dec1969, PDF p. 62

NOTE: The legend included in Figure 5-27 applies to Figure 5-28 as well.

Figure 5-28: EBR-II Reactor Building (top view by levels)
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EBR-II Missions

Beginning in 1964 and throughout its remaining history, EBR-II had several overlapping operating 

missions (Sackett, undated). The reactor’s first mission effectively ended in April 1969 when the last 

recycled fuel subassembly was loaded into the reactor (Sackett, undated; EBR-II, undated). The latter 

three missions continued beyond 1979 (the end date of the period under evaluation) (Sackett, 

undated). Figure 5-29 portrays EBR-II’s four operating missions, followed by a summary of each 

mission.

Source: Sackett, 2010, derived from figure on PDF p. 15

Figure 5-29: The EBR-II Reactor’s Four Operating Missions

Mission I: EBR-II, along with its support facilities (Sodium Boiler Plant, Power Plant, and Fuel 

Cycle Facility), were originally built to demonstrate the feasibility of on-site fuel reprocessing as 

an adjunct to a liquid-metal-cooled, fast-breeder-reactor power plant (Stacy, 2000; EBR-II Fact 

Sheet and Memo, 1968). Together, the EBR-II reactor and the Fuel Cycle Facility formed a 

closed fuel-cycle operation (i.e., spent fuel from the reactor was transferred to the Fuel Cycle 

Facility for pyrochemical re-processing, re-enrichment, and re-fabrication, and then it was 

returned to the reactor) (EBR-II Fact Sheet and Memo, 1968). During the reactor’s initial 

mission, the primary emphasis was on verifying and demonstrating the operating characteristics 

of the power cycle and the fuel cycle (EBR-II, undated). Particular attention was given to the 

feasibility of the unique EBR-II fuel cycle, especially the onsite processing and fabrication of 

highly-radioactive fuel and the power performance of recycled fuel (EBR-II, undated). The 

objectives of that initial mission were met within the first few years of EBR-II’s operation 

(Stacy, 2000). 
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Mission II: EBR-II’s second mission began in May 1965 when the first irradiation experiment 

was installed in the reactor’s core (EBR-II, undated). In 1969, the experimental focus was 

reoriented from a demonstration plant to an irradiation facility for candidate LMFBR fuels and 

materials (EBR-II, undated; OSTI 6345424). In September 1970, some of the reactor’s depleteduranium blanket subassemblies were replaced with stainless-steel reflector subassemblies to 

enhance the reactor’s irradiation capabilities for its fuel-testing and materials-development work 

(Sackett, undated; Case, 1971). During the second mission, a wide variety of uranium, 

plutonium, and thorium reactor fuels were tested, including metal, oxide, carbide and nitride 

fuels (Sackett, undated). Those tests also helped develop the reactor fuels for both the Fast Flux 

Test Facility and the Clinch River Breeder Reactor Plant (Sackett, undated). It was from that 

experience that the benefits of metal fuel irradiation performance were established (Sackett, 

undated). In the late-1970s, the breeding part of the second mission was dropped and the 

mission was modified to include irradiating candidate fuels and materials for all types of Liquid 

Metal Reactors (LMRs). 

Mission III: In the late-1970s, EBR-II’s third mission began. It expanded the effort from a 

conservative steady-state fuels testing program to an operational reliability testing (ORT) 

program (Sackett, undated; Lentz, undated). The ORT program used EBR-II for a more 

aggressive irradiation program consisting of: (1) run-beyond-cladding-breach (RBCB) tests; (2) 

operational transient tests simulating duty-cycle transients and mild over-power transients on 

fuel elements; and (3) thermal-hydraulic testing of clusters of fuel elements under both normal 

and natural convection modes of cooling for testing and verification of shutdown heat-removal 

codes (Lentz, undated). For the reactor transient tests, plant upgrades included providing special 

instrumented test facilities for in-core measurements, fission-gas-handling systems for breached 

fuel testing, and computer-controlled power-shaping for transient testing (Sackett, undated). The 

most dramatic result of this work was the series of tests conducted in 1985–1986 that culminated 

in a demonstration of the reactor’s ability to accommodate loss-of-flow and loss-of-heat-sink 

transients without needing to scram the reactor (Sackett, undated). 

Mission IV: In the mid-1980s, EBR-II’s fourth mission involved fuel development and testing 

activities for the Integral Fast Reactor (IFR) Program (Lentz, undated). No further details are 

provided in this document regarding these activities because all of them occurred after the period 

under evaluation. 

EBR-II Reactor Description

EBR-II was an unmoderated, sodium-cooled, fast-neutron power reactor, which was designed to 

produce 62.5 megawatts of heat (62.5 MWt) and 20 megawatts of electricity (20 MWe). EBR-II is a 

piped-pool design with the entire primary system and reactor vessel submerged in a large pool of 

sodium contained in the primary tank (Sackett, undated). Heat produced in the reactor is removed by 

the primary sodium system and transferred to the secondary sodium system by the heat exchanger. 

From the secondary system, the heat is transferred in the steam generator to produce superheated 

steam, which is then delivered to a conventional condensing turbine at 850 oF and 1250 psi (EBR-II 

Fact Sheet and Memo, 1968; Lentz, undated).
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Until September 1970, EBR-II consisted of a high-enriched uranium core surrounded on all sides by a 

blanket of depleted uranium (EBR-II Fact Sheet and Memo, 1968; Sackett, undated; Case, 1971). 

While producing power, the reactor was designed to manufacture (breed) plutonium, another fuel, 

from the U-238 in the blanket material. Furthermore, the reactor was designed to breed more fuel 

than it consumed. The net effect of this breeding process is to extend many-fold the United States’ 

natural and valuable supplies of uranium (EBR-II Fact Sheet and Memo, 1968). In September 1970, 

some of the EBR-II Reactor’s depleted-uranium blanket subassemblies were replaced with stainlesssteel reflector subassemblies (Sackett, undated; Case, 1971). 

EBR-II Timeline Through 1979 

Table 5-2 summarizes the important dates in EBR-II’s operating history from 1957 through 1979. 

Table 5-2: EBR-II Timeline (1957-1979) 

Date EBR-II Activity

December 19, 1957 Construction of the EBR-II’s containment vessel started (EBR-II, undated; Smith, 1977).

November 1960 Construction of EBR-II’s power plant and reactor plant was completed (EBR-II, undated).

April 1961 Fabrication of the fuel for the first core completed in Chicago, Illinois at ANL-E Site (EBRII, undated; Chakraborty, 1971). 

September 30, 1961 Dry criticality (i.e., low-power operation without sodium coolant) of reactor achieved 

(EBR-II, undated).

October 30, 1963 Loading of the reactor for the approach to wet critical was started with the transfer of 17 

enriched uranium subassemblies from the storage basket to the reactor grid (Progress 

Report, 1963, PDF p. 16; Kirn, 1964).

November 11, 1963 Wet criticality (i.e., with sodium coolant) of reactor achieved (EBR-II, undated; Progress 

Report, 1963, PDF p. 16; Kirn, 1964).

December 1963 – May 

22, 1964

Reactor did not operate due to repairs being completed on both primary sodium pumps and 

some reconfiguration of the coolant system. (Progress Report, May1964, PDF p. 52; 

Familiarization Effort, 1970)

July 16, 1964 Approach-to-power started. Approval was received to begin the approach-to-power 

experiments on July 16, 1964, and the reactor was made critical at 20:30 hours that day.

(Progress Report, Jul1964, PDF p. 41; EBR-II, undated; Perry, 1978)

September 1964 First irradiated (0.1% burnup) fuel was removed from EBR-II for examination and 

reprocessing (EBR-II Fuel Cycle Story, 1987, PDF p 248).

May 1965 The first recycled fuel subassembly (subassembly C162) was made at the FCF in April 1965 

and installed in the reactor core in May 1965 (EBR-II, undated; EBR-II Fuel Cycle Story, 

1987, PDF p 248; Progress Report, Dec1965, PDF p. 18). 

May 10, 1965 EBR-II began its role as an irradiation facility when the first two experimental 

subassemblies were inserted into the reactor core for irradiation experiments during Run 5. 

One experimental subassembly contained mixed oxides for an experiment sponsored by 

General Electric, and the other subassembly contained U-Pu-Fissium alloy for an 

experiment sponsored by ANL’s Metallurgy Division. (Staker, 1975; EBR-II, undated; 

Chakraborty, 1971, PDF p.15; Smith, 1977)
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Date EBR-II Activity

May 24, 1967 The first fuel assembly failure occurred in an experimental fuel assembly, which was 

signaled by a release of fission gases (Familiarization Effort, 1970; EBR-II Fuel Cycle 

Story, 1987, PDF p. 229; Progress Report, May1967, PDF pp. 15-17). This resulted in 

radioactivity being released into occupied areas of the reactor building. The six workers 

that were in the reactor building during the release event received whole body counts, but 

no intakes of radioactive material were detected. Xe-133, Xe-135, Xe-138, and Cs-138 

were detected in samples of the reactor’s cover gas. (Progress Report, May1967, PDF pp. 

15-17)

June 10, 1967 A small radioactive sodium fire occurred in the primary sodium sampling room. Thereafter, 

access to the sampling room was restricted until the Na-24 radioactivity had decayed. Some 

Na-24 was transported out of the sampling room by the ventilation system, and caused 

minor air and surface contamination in the reactor building. (Progress Report, Jun1967, 

PDF pp. 47-48)

April 1969 Last recycled fuel assembly installed in reactor core (EBR-II, undated).

Mid-November 1970 –

March 1971

Reactor was shut down for about four months between Runs 47A and 47B (Chakraborty, 

1971, PDF p. 22).

1972 The first of the In-Core Test (INCOT) Facilities installed in reactor core. The INCOT 

Facilities provided direct access into the reactor core to accomplish operational testing of 

in-core instrumentation and materials under simulated LMFBR conditions. (Staker, 1975) 

Also, some of the experimental fuel elements were tagged with “xenon tags,” which were to

be used to identify the fuel element that failed (Smith, 1977). 

April-June 1972 Reactor was shut down for about three months between Runs 55B and 56A (Case, 1974).

June 20, 1977 The first Run-Beyond-Cladding-Breach (RBCB) experiment was irradiated. The 

experiment contained a pre-irradiated mixed oxide fuel element with a machined slit in its 

cladding. (Smith, 1977) 

5.1.2.3 Fuel Cycle Facility (FCF)

ATTRIBUTION: Section 5.1.2.3 was completed by Brian Gleckler, Dade Moeller, Inc. All 

conclusions drawn from the data regarding FCF were peer-reviewed by the individuals listed on the 

cover page. The rationale for all conclusions in subsequent sections of this document devoted to FCF 

are explained in the associated text.

The FCF performed radiological activities at various capacities from July 1963 through 1979 and 

beyond. The FCF was constructed as a pyrometallurgical reprocessing plant to meet EBR-II’s nuclear 

fuel needs by reprocessing the breeder reactor’s driver and blanket fuels. The intent was to reprocess

the irradiated depleted-uranium blanket fuels to extract the plutonium that had been bred. However, 

the equipment for this reprocessing was never installed. 

The various steps to the pyrometallurgical reprocessing process took place in the FCF’s two hot cells 

(i.e., the Air Cell and Argon Cell), and this irradiated fuel production line was later referred to as the 

FCF’s “Hot-Line.” The Hot-Line included processes for disassembling, refining, refabricating, and 

reassembling EBR-II fuel subassemblies. In December 1966, a program was announced that would 

significantly increase EBR-II’s fuel production in order to meet the increasing need for EBR-II reactor 

fuel. This program included plans to procure reactor fuel from off-site commercial sources, and the 
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on-site fabrication of unirradiated fuel elements outside of the FCF’s hot cells (i.e., Cold-Line fuel 

fabrication). This resulted in Cold-Line fuel fabrication processes being set up in the outer rooms of 

the FCF’s main (ground) floor (i.e., those rooms surrounding the Argon Cell). In October 1968, an 

announcement of programmatic changes for the FCF’s Air and Argon Cells was received (Progress 

Report, Oct1968). Those changes included a gradual decrease in Hot-Line fuel production and an 

increase in support for the experimental irradiation programs (Progress Report, Oct1968). By March 

1969, the FCF’s Hot-Line was phased out and the Air and Argon Cells would only be used for 

shipping, inspecting, and testing irradiated reactor fuels (Progress Report, Mar1969; Progress Report,

Oct1968). 

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In Figure 5-24, the FCF is Building 765. The FCF is also known by the following building numbers: 

ANL-765, EBR-II-602, and MFC-765. Maps of the various areas in the FCF are depicted in Figures 

5-30 through 5-32. 

The FCF has had a number of name changes to reflect changes in the type of work being performed at 

the facility. Between 1964 and 1969, when a demonstration of technology for recycling nuclear 

reactor fuel was being performed, the facility was called the Fuel Cycle Facility. Between 1970 and 

1971, the facility was sometimes referred to as the Fuels and Examination Facility (FEF) (Progress 

Report, 1971, PDF p. 22; Cushman, 1970, PDF p. 32). The facility was often referred to as the FCF 

and FEF in the same documents. In various 1972 documents, the FCF was often referred to as the 

FEF Complex, and then the HFEF Complex, even though the other main part of that complex 

(HFEF-N) was still under construction. By the time HFEF-N construction was completed in 1972, the 

FCF was renamed the Hot Fuel Examination Facility-South (HFEF-S and/or HFEF/S) (Operations 

Training, 2009; Site Plan, 1981). In 1993, the FCF was renamed the Fuel Conditioning Facility 

(Operations Training, 2009). 

FCF Missions

FCF Mission I: Between 1964 and 1969, the FCF was operated as a demonstration of Liquid 

Metal Fast Breeder Reactor (LMFBR) technology for the remote reprocessing and re-fabrication 

of nuclear reactor fuel from a closed cycle with the EBR-II reactor (HFEF History, 1979; 

Stevenson, 1987). This irradiated fuel process was referred to at the FCF as Hot-Line fuel 

production. By 1969, the FCF had successfully completed its mission of demonstrating the 

processing and recycling of spent fuel to support EBR-II reactor operations (Courtney, 1991). 

FCF Mission II: In October 1968, an announcement of programmatic changes for the FCF’s Air 

and Argon Cells was received (Progress Report, Oct1968). Those changes included a gradual 

decrease in Hot-Line fuel production and an increase in support on the experimental irradiation 

programs (Progress Report, Oct1968). By March 1969, the FCF’s Hot-Line was phased out, and 

the Air and Argon Cells would only be used for shipping, inspecting, and testing irradiated 

reactor fuels (Progress Report, Mar1969; Progress Report, Oct1968). In support of EBR-II 

reactor operations, the FCF handled, packaged, and shipped spent fuels to the Idaho Chemical 

Processing Plant (ICPP) for reprocessing, and packaged blanket and reflector subassemblies for 

disposal (HFEF History, 1979). The FCF also provided hot cell services for the qualification of 

EBR-II reactor driver fuel, blanket, and reflector subassemblies, and for handling and examining 

irradiated components (HFEF History, 1979). The FCF also provided interim and 

post-irradiation handling, along with destructive and nondestructive examinations of fuel and 
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material experiments from the EBR-II and TREAT reactor irradiation programs (HFEF History, 

1979; 111648). 

Source: Environmental Assessment, 1996, derived from figure on PDF p. 26

Figure 5-30: General Layout of FCF’s Main Floor (ground floor)
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Source: Hesson, 1963, derived from figure on PDF p. 52

Figure 5-31: Original Layout of FCF’s Basement
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Source: Courtney, 1991, PDF p. 5

Figure 5-32: Elevation Cross-Section of FCF
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FCF Mission III: Starting in late-1969, the FCF began supporting RAS-TREAT sodium-loop 

experiment work, which was an intermittent activity that was initially performed in the FCF’s 

Sodium Cleanup Room (Room 27) (FCF Rad Survey, 1970, PDF pp. 208-211; 145251). The 

RAS-TREAT sodium loop experiments were sponsored by ANL-E’s Reactor Analysis and 

Safety (RAS) Division, and were irradiated in the TREAT Reactor (Cook, 1975). Because the 

work performed in FCF Rooms 26–29 (and eventually Room 22) was outside of the hot cells, 

only unirradiated and slightly-irradiated RAS-TREAT sodium loops were initially handled at the 

FCF. Initially, the FCF’s RAS-TREAT sodium-loop experiment work only involved the Mark-I 

type loops that only contained uranium test fuels, and only involved draining and washing the 

NaK alloy metal coolant from the loops. However, in 1970, the RAS-TREAT sodium loop 

experiment work at the FCF started to include the Mark-II type loops, which could contain 

uranium-plutonium and possibly plutonium test fuels (FCF Rad Survey, 1970, PDF pp. 208-

211). 

An April 1970 survey is the first indication of Mark-II loop work taking place at the FCF, which 

was performed in the FCF’s Mock-Up Area (Survey Reports, Apr1970). However, this initial 

Mark-II loop work only involved draining and washing the NaK alloy metal coolant from the 

loops. By April 1970, Mark-II loop work at the FCF started to include the removal of the test 

sections from the Mark-II loops in the Sodium Cleanup Room (Survey Reports, Apr1970, PDF

pp. 36-48). The start of the potential exposures at the FCF to plutonium without mixed fission 

products being present would have begun with removing and handling the test sections from the 

Mark-II loops. In early-1971, that work was moved to a newly-installed inert-atmosphere 

glovebox in Room 22, which was the former FCF Machine Shop (Progress Report,

AprMay1971). In early-1973, the FCF’s Argon Cell was decontaminated in preparation for 

handling and processing more highly-irradiated sodium-loop experiments (Cook, 1975). During 

1973 and 1974, the first Mark-IIA sodium-loop experiments were sectioned in the Argon Cell. 

The purpose of the sectioning was to reduce the sodium-loop experiments into smaller segments 

for shipment to the Materials and Science Division (MSD) at ANL-E (Cook, 1975).

Background: The integral sodium loop is a package unit in which fast-reactor fuel pins can be 

tested to destruction in stagnant or flowing sodium coolant while being irradiated 

within different types of reactors (e.g., air-cooled and water-cooled). The loops 

contained a sodium reserve and flow system, electromagnetic pump, power 

transformer, sodium flowmeter, sodium heaters, expansion chamber, fuel-test 

section, pressure and temperature instrumentation, along with power and control 

wiring. The sodium loops fit into either a standard single- or double-size dummy 

fuel assembly for the reactor in which they were to be irradiated (Annual Report, 

1965; Robinson, 1971).
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FCF Mission IV: In 1990, it was proposed that the FCF be modified to once again reprocess 

EBR-II reactor fuel as part of a test for the Integral Fast Reactor (IFR) concept. The ultimate 

purpose of the proposed action was to provide essentially-complete fuel-cycle service for the 

EBR-II reactor. The proposed fuel cycle program was needed to establish the feasibility of the 

IFR concept, which includes proving the feasibility of a high-temperature-metal fuel cycle for 

advanced liquid-metal-cooled reactors. The IFR program was considered important to the 

national interest because of its potential to improve nuclear power economics and because of the 

inherent safety achievable through use of liquid-metal-cooling and metallic fuels. Other 

potential benefits of this fuel-cycle included proliferation resistance of the processed fuel 

because of residual inherent radioactivity (DOE, 1990). No further details are provided in this 

document regarding this mission, because it occurred after the evaluation period for this report. 

FCF Facility Description

The FCF consists of four main operating areas: Reactor Passageway, Air Cell, Argon Cell, and the 

Auxiliary Areas. The Auxiliary Areas are further subdivided into three main areas: the Operating 

Floor Auxiliary Areas, Service Floor Auxiliary Areas, and Roof Area. (Hesson, 1963; Stevenson, 

1987) 

Over time, manipulator and slave decontamination, maintenance, and repair became a major activity 

performed throughout the FCF. Based on the available survey records, decontamination work and 

manipulator and slave repair work was at times performed in nearly every room of the FCF’s 

Auxiliary Areas. This included some rooms originally designated as office areas. As fuel 

reprocessing came to an end at the FCF, more and more of the offices areas were converted over to 

work areas and production areas for new projects and activities. 

Reactor Passageway

The Reactor Passageway was the passageway between the FCF and the EBR-II reactor. It was used to 

transfer reactor fuel and experiments (irradiated and unirradiated) between the FCF and the EBR-II 

reactor. Those reactor fuels and experiments were transferred via an Inter-Building Coffin (IBC). 

The Reactor Passageway also served as a sodium decontamination area for materials being removed 

from the EBR-II reactor. A description of the activities performed in the Reactor Passageway is

provided later in this section in Hot-Line Step 1. The pathway that the IBCs would travel between 

EBR-II and the FCF is shown in Figure 5-33. 
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Source: Forrester, undated

Figure 5-33: Elevation Cross Section of Reactor Passageway

Air Cell

The FCF’s Air Cell is a large (15 ft. wide x 47ft long x 21 ft. high) hot cell with an air atmosphere. 

The original primary purpose of the Air Cell was to provide a radiation-shielded area with an air 

atmosphere where reactor subassemblies could be disassembled and reassembled, and where fuel 

elements could be inspected and tested by remote methods. Initially, the Air Cell was the only 

terminal for the air locks to the Argon Cell. The Air Cell also served as a place for servicing Argon 

Cell equipment and for preparing scrap for disposal. The Air Cell is constructed of reinforced 

high-density (3.5 g/cm3) barite concrete. The east end of the Air Cell shares a wall with the Argon 

Cell. The Air Cell is equipped with nine viewing windows, one periscope opening, one optical 

peephole sleeve in the walls, a roof hatchway with stepped removable plugs served by a jib crane, and 

two plugged viewing positions in the roof. The jib crane was later removed and the roof hatch was 

modified. Each window position is provided with two sleeves for master/slave manipulators and with 
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sleeves through the cell floor to the basement for service leads. The Air Cell is also equipped with 

one five-ton crane and two bridge-type manipulators. The bridge manipulator operates on rails along 

the cell walls below the crane’s rails. Processing equipment inside the Air Cell is limited to 10 ft. in 

height due to the elevation of the bridge manipulators (Hesson, 1963; Stevenson, 1987). 

The Air Cell is also equipped with a two-part entry door for personnel, which was normally sealed, 

one large transfer port to the basement (a.k.a. Service Floor), and one small transfer port to the 

operating area. Irradiated fuel subassemblies from the EBR-II reactor were brought into the Air Cell 

via the large transfer port. There are also one large and two small transfer locks to the Argon Cell. 

The two smaller transfer locks were used to transfer small items (e.g., fuel elements, fuel cans, 

crucibles, fume traps, samples) directly into the Argon Cell. Larger materials and equipment were 

transferred through the Air Cell’s large (6-ft diameter) transfer lock to the Argon Cell via the Transfer 

Cell, which is located below the Air and Argon Cells. The Transfer Cell (a.k.a. Subcell 4) served at 

the air lock between the Air and Argon Cells (Hesson, 1963; Stevenson, 1987). 

The Air Cell is maintained at a negative pressure relative to the occupied areas in the FCF, and is 

ventilated at the rate of 5000 ft3

/min. Building air from outside the Air Cell enters the cell through 

dampers at the upper part of the cell wall at the west end and is exhausted through the Transfer Cell, 

which is below the east end of the cell floor. This creates a downward airflow throughout the cell, 

which minimized the spread of contamination. After passing through a bank of fine-particle filters, 

the air leaving the Air Cell is released to the atmosphere through a 200-ft-tall stack (see Figure 5-34) 

(Stevenson, 1987).

Argon Cell 

The FCF’s Argon Cell is an annular-shaped hot cell with an inert-argon atmosphere. The original 

primary purpose of the Argon Cell was to provide a radiation-shielded area with an inert-argon 

atmosphere where metallic reactor fuels and sodium can be safely exposed (Hesson, 1963; Stevenson, 

1987).

The Argon Cell is in the shape of a 16-sided regular polygon that has a central control area (see Figure 

5-30). The west side of the Argon Cell shares a wall with the Air Cell, each of the other 15 sides is 

provided with a shielded window or viewing device. The central control area is also in the shape of a 

16-sided regular polygon, but with only eight shielded windows (one window in alternate faces). The 

Argon Cell is 22 ft. high, has a 16-ft-wide annular area between the two polygons with a total floor 

area of 2,000 ft2 and a total volume of about 60,000 ft3 (Stevenson, 1987). The walls on the operating 

level of the Argon Cell are made of reinforced high-density (3.5 g/cm3) barite concrete (Hesson, 1963; 

Stevenson, 1987). The Argon Cell is also completely lined with zinc-coated sheet steel (Stevenson, 

1987). 

On the west end of the Argon Cell, there are two small transfer locks to the Air Cell in the wall 

between the cells, and one large (6-ft diameter) transfer lock in the floor, which also leads to the Air 

Cell via the Transfer Cell. There are also six viewing plugs, a plugged-and-sealed overhead 

hatchway, and three periscope openings in the roof (Stevenson, 1987).

The cell is provided with two five-ton cranes and six manipulators (Hesson, 1963; Stevenson, 1987). 

The crane and manipulator bridges pivot in the center of the cell above the central control area, and 

the outer ends operate on circular rails. The crane bridges operate on outer rails 17 ft. above the floor 
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and on inner rails suspended from the roof. The manipulators operate on outer rails 11.5 ft. above the 

floor and on inner rails mounted on the central control area roof. Process equipment height is limited 

to 10 ft. The cranes cannot pass each other but can pass above the manipulators. The manipulators 

cannot pass each other (Hesson, 1963).

There is a suspect stack. Its purpose is to provide a high point of discharge to the atmosphere of all 

ventilation atmosphere suspected of containing radioactive contamination. There is a hold-up tank to 

provide a place for storing active gases and vapors from the melt-refining operations (and possibly 

skull oxidation) to await favorable meteorological conditions for stack release. As a secondary effect, 

some decay of activity can take place in the hold-up tank. Xe-133 and possibly I-131 were the chief 

radionuclides to undergo decay (Hesson, 1963).

The ventilation air is filtered before being discharged to the stack. Gases from process operations are 

filtered through high-efficiency filters and, in most cases, are further processed (e.g., through a tower 

for iodine removal) before being released to the stack. Each of the locations has booster fans and 

filters. The FCF has two 38,000-cfm main booster fans. The air flow in the ductwork, filters, and 

piping is controlled at a negative pressure. 

Operating Level Auxiliary Areas

The major radiological work areas on the Operating Level are described below.

Mock-Up Area: The Mock-Up Area was initially an area in which the feasibility of proposed 

remote operations could be determined (Hesson, 1963). Later, the area was also used to test and 

develop equipment for cladding the fuel pins and testing the completed elements (Stevenson, 

1987). Eventually, the area was used for a multitude of activities, some involving highly 

radioactive equipment and materials (e.g., manipulator repairs. 

Room 25: This is the Mold Preparation Room, and was sometimes referred to as the Thoria 

Room. Thoria (ThO2) was used to coat the molds until Cold-Line (unirradiated) fuel production 

operations began in July 1967. After Cold-Line operations began, zirconium oxide was used 

instead of thoria to coat the molds (Progress Report, Dec1967). As early as January 1967, 

survey records indicate that depleted uranium (DU) fuel pins were being sent to this room for 

cleaning prior to being sent elsewhere for radiography. Survey records also indicate that 

enriched uranium was being handled in this room and that the room was also being used for 

manipulator decontamination and repairs. 

Room 20: Until 1967, Rooms 19 and 20 were just office areas. By August 1967, Room 19 was 

incorporated into Room 20 to create the Fuel Pin Manufacturing Facility (Batte, 1986; Stoddart, 

1967). The reconfigured Room 20 was the initial and primary location for the Cold-Line for 

EBR-II fuel production for several years (i.e., the unirradiated fuel production line). Operations 

in Room 20 included: (1) casting fuel pins; (2) shearing fuel pins to length; (3) removing glass 

mold material; and (4) dimensionally inspecting the fuel pins prior to encapsulating them into 

elements and subassemblies (Batte, 1986; Stoddart, 1967). The first injection casting run was 

made in Room 20 on August 26, 1967 (Stoddart, 1967).

At the end of December 1969, the Cold-Line fuel pin and fuel element production equipment in 

FCF Rooms 20 and 26 was put on stand-by (Progress Report, Dec1969, PDF p. 88). In the mid-
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1970s, the FCF’s Cold-Line fuel pin and fuel element production equipment was deactivated. 

Fuel pin fabrication in the FCF was not reinitiated until 1982 (Batte, 1986, PDF p. 9).

Room 26: Prior to July 1967, Room 26 was referred to as the Degas Room, Measurements 

Laboratory, and Precision Measurements Lab, and contained a vacuum-type induction furnace, 

air gauges, an optical comparator, a polariscope, and various components for equipment in the 

hot cells (i.e., in-cell equipment). During that period, Room 26 appears to have been used only 

for: (1) melt-refining-crucible preparation; (2) cladding and Vycor mold inspection; and (3) 

calibrating and testing of in-cell equipment (Hesson, 1963).

By July 1967, Room 26 had been reconfigured to be part of the FCF’s Cold-Line. The available 

information indicates that Room 26 was used for preparing the uranium-fissium alloy for 

injection-casting in Room 20 and removing the skull (dross) from the casting crucibles. As a 

result, this room was sometimes referred to as the Alloy Prep Room. In July 1967, five test runs 

on the alloy-preparation furnace (a.k.a. melt refine furnace) were completed using chopped-up 

depleted-uranium-fissium alloy pins (Progress Report, Jul1967). The first enriched uranium 

ingot for the Cold-Line was melted and poured in Room 26 on August 17, 1967 (Stoddart, 1967).

Room 23: Initially, Room 23 was referred to as the Glove Box Laboratory or Dry Boxes Room, 

and contained a sodium extruder, a degreasing tank, and low-moisture, low-oxygen argon 

gloveboxes (dry boxes) (Hesson, 1963). At that time, Room 23 appears to have been used only 

for: (1) sodium extrusion and loading into fuel element cans; and (2) performing maintenance 

and service of processing equipment (Hesson, 1963). Survey data indicates that depleted and 

possibly enriched uranium were handled in Room 23. 

After August 1967, Room 23 was used for some of the Cold-Line fuel production activities, but 

it is uncertain what was done in this room. Given that one of the gloveboxes in Room 23 was 

converted into the “Cold-Line glovebox,” that glovebox was likely used to assemble the 

Cold-Line fuel elements and perform the sodium-bonding step in the process. 

Room 27: Room 27 was a steel-lined room with a blowout wall and sodium-disposal hood 

(Hesson, 1963). The room was used for sodium disposal and for decontaminating various items 

(sodium and radiological contaminants) (Hesson, 1963). The room was also known as the 

Sodium Cleanup Room, Sodium Disposal Room, and Decontamination Room. Prior to 1970, 

potential actinide exposures without fission products present were limited to depleted and 

enriched uranium. 

Based on the survey records, work with the Mark-I TREAT Sodium Loops in Room 27 began in 

February 1970 (Survey Reports, Feb1970, PDF pp. 81-85; FCF Rad Survey, 1970, PDF p. 208). 

However, this initial TREAT Sodium Loop work only involved loops containing metallic 

uranium fuel. The work was performed under Non-Routine FCF Operation Proposal No. 1-391 

(a.k.a. NR1-391) (Survey Reports, Feb1970, PDF pp. 81-85). 

Later in 1970, work involving the disassembly of Mark-II TREAT Sodium Loops, which 

sometimes contained “PuO and UO2

”

, began in this room (FCF Rad Survey, 1970, PDF pp. 208-

211). This work was performed under Non-Routine FCF Operation Proposal No. 1-398 (a.k.a. 

NR1-398) (Survey Reports, Apr1970, PDF pp. 36-48). Based on the Propose to Carry Out date 

on NR1-398, this activity did not begin until April 14, 1970 (Survey Reports, Apr1970, PDF pp. 
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36-48). Based on the survey and air monitoring records, this Room 27 work was sporadic and 

periodically resulted in some alpha contamination (surface and airborne). However, the Health 

Physics Instructions for NR1-398 required that, upon job completion, the area be surveyed and 

that the area and equipment be decontaminated, as necessary (Survey Reports, Apr1970, PDF pp. 

36-48). This activity appears to have taken place in Room 27 until around April 1971 (Progress 

Report, AprMay1971). Based on the available information, a glovebox was installed in Room 

22 for this activity and this work was transferred to Room 22 after the glovebox became 

operational (Progress Report, AprMay1971). 

Room 22: This room was initially the FCF Machine Shop until sometime between 1968 and 

1971. In 1968, construction was completed on Building 782, which was built to serve as a 

centralized Machine Shop and receiving inspection facility for the EBR-II Area (Site Plan, 

1981). A November 4, 1970 survey is the first instance found where Room 22 was referred to as 

the “Old Machine Shop.” As early as January 1971, Reactor Development Program Progress 

Reports were indicating that a glovebox for handling the Mark-II TREAT Sodium Loop was 

being installed in Room 22. While Room 22 operated as the FCF Machine Shop, survey records 

indicate that materials containing depleted and enriched uranium were machined there, and that 

some of those materials were irradiated.

Sometime after May 1971, the argon atmosphere glovebox in Room 22 became operational and 

work with the Mark-II TREAT Sodium Loops started (Progress Report, AprMay1971). A 

survey record from August 17, 1972 is the first indication of Pu in the Room 22 glovebox, which 

is also when they started collecting much more air monitoring data for Room 22. 

Room 30: This was a small triangular-shaped room that was equipped with a sink and was used 

to decontaminate small materials and equipment. The room was called the Small 

Decontamination Room (Health Physics Monthly Report, Jan1968, PDF p. 32), but it is 

suspected that it may have also been the room that is periodically referred to as the “Janitor’s 

Closet” in a number of the survey records. Survey records indicate that this room was normally 

maintained as a contamination area.

Room 12: This was designated as an office. Starting in April 1970, survey records and 

Non-Routine FCF Operation Proposal No. 1-400 (NR1-400) indicate that high-pressure rupture 

tests were being performed on irradiated tubing using a BRATT furnace in Room 12. NR1-400 

also indicated that no fuel was involved with those tests, and the primary radionuclides of 

concern were Co-58 and Mn-54. (Survey Reports, Apr1970, PDF pp. 74-76).

Service Floor Auxiliary Areas

The primary feature of the FCF Service Floor (a.k.a. Basement) was the Sub-Cells, which were 

directly below the Air and Argon Cells. The Sub-Cells were areas where the auxiliary equipment for 

the Air and Argon Cells was maintained. In December 1968, modifications to Sub-Cell 4 (a.k.a. 

Transfer Cell), which was used to transfer large items between the Air Cell and Argon Cell, were 

completed. These modifications allowed items (mostly manipulators) to be transferred directly 

between Sub-Cell 4 and the basement (Progress Report, Dec1968, PDF p. 72). This also eliminated 

the need for all large items to pass through the Air Cell before going into the Argon Cell. In late 
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1970, a manipulator decontamination and repair area was constructed in the basement (Survey 

Reports, Oct1970, PDF p. 63). Figure 5-34 shows the New Hot Repair Area. 

Source: Courtney, 1991, derived from figure on PDF p. 8

Figure 5-34: New Hot Repair Area in FCF’s Basement

Roof Area

Directly above the ceiling hatches to the Air Cell, the FCF’s Roof Area (a.k.a. the Penthouse) was 

used as a radioactive equipment repair area (mostly for manipulators), and initially was completely 

exposed to the outside environment. In time, a “Wind and Weather Enclosure” was built around this 

area on the roof. In May 1968, construction of a full-blown manipulator decontamination and repair 

facility was started (Progress Report, May1968, PDF pp. 81, 91). This facility would provide a place 

to decontaminate in-cell cranes, manipulators, and other large pieces of equipment prior to attempting 

their contact maintenance. Construction of this roof-top facility was not completed until sometime 

after December 1968 (Progress Report, Dec1968, PDF p. 72). 
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Overview of the FCF Hot-Line Fuel Production Process 

The FCF used a pyrometallurgical process to reprocess EBR-II reactor fuel into usable fuel. Hot-Line 

(irradiated) fuel production involved 14 major processing steps that occurred almost entirely inside the 

FCF’s Air and Argon Cells. The primary feed for the Hot-Line was irradiated fuel subassemblies 

(a.k.a. driver subassemblies) from the EBR-II reactor. To a much lesser extent, fresh (unirradiated) 

fuel was also fed into the process as make-up fuel, to replace the U-235 that was burnt up in the 

reactor. 

The Hot-Line’s 14 processing steps are summarized below (Hesson, 1963; Koch, undated; Stevenson, 

1987; Stevenson, 1969; Fuel Cycle Facility, undated):

Hot-Line Step 1 – Transfer of Subassemblies to FCF & Sodium Removal: After a 14-day cooling 

period, irradiated fuel subassemblies were ready to be transferred from the EBR-II reactor to the 

FCF. At the reactor, an individual subassembly was transferred from the reactor’s subassembly 

storage rack into a shielded portable cask called an Inter-Building Coffin (IBC). The IBCs were 

transferred from the EBR-II Reactor to the FCF’s Reactor Passageway on a cart. While in the 

Reactor Passageway, a simple procedure was performed to remove the residual sodium on the 

subassemblies while still in the IBC. The transfer of the IBCs from the EBR-II Reactor to the 

FCF’s Air Cell is depicted in Figure 5-33 (Stevenson, 1987, PDF pp. 259-260; Fuel Cycle 

Facility, undated, PDF p. 8).

Hot-Line Step 2 – Subassembly Disassembly: In the FCF’s Air Cell, the irradiated fuel 

subassemblies were received and remotely-disassembled into individual fuel elements. The fuel 

elements were then loaded into magazines and then passed through one of the small transfer 

locks into the Argon Cell (Hesson, 1963; Koch, undated; Fuel Cycle Facility, undated).

Hot-Line Step 3 – De-canning & Pin Chopping: After the irradiated fuel elements were received 

in the Argon Cell, the fuel elements went through the de-canning process (Koch, undated; 

Stevenson, 1969; Fuel Cycle Facility, undated). The de-canning process reduced the fuel 

elements to bare fuel pins, and mechanically removed the cladding from the fuel (Koch, undated; 

Stevenson, 1969; Fuel Cycle Facility, undated). Prior to a melt-refining run, approximately 150 

irradiated fuel elements were selected for a given melt-refining run (Stevenson, 1987, PDF p. 

104). The selected fuel pins with their adherent sodium were then chopped into approximately 

1.5-inch sections by a mechanical chopper (Koch, undated; Stevenson, 1969; Fuel Cycle Facility, 

undated). The chopped fuel pins were then transferred to a pin charger before going to the Melt 

Refining Station (Stevenson, 1987, PDF p. 105). 

Hot-Line Step 4 – Melt Refining: Melt refining is also known as "oxidative slagging" or 

"drossing.” As indicated above, approximately 150 irradiated fuel elements were selected for a 

given melt-refining run. Outside the Argon Cell, the required quantities of the alloying 

constituents (U-235, U-238, fissium metals, a master silicon alloy, if required) were weighed out 

from unirradiated stocks and added to a new zirconia (ZrO2) crucible. The partially-charged 

ZrO2 crucible was then transferred into the Argon Cell and placed in the melt-refining furnace. 

Next, the pin charger that came from the pin chopper was positioned over the ZrO2 crucible and 

tipped by the in-cell manipulator, allowing the pin charge to slide into the crucible. With the 

crucible’s charge completed, the Fiberfrax® fume trap was then placed over the crucible and the 
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bell jar was placed over the furnace and sealed. These runs typically involved 10–12 kg charges, 

which yielded an average ingot size of 10 kg. The charge in the crucible was then melted and 

poured into a graphite ingot mold with a pyrolytic carbon coating (Stevenson, 1987, PDF pp. 

95-129, 259-260). 

After the melt-refining furnace had cooled down, the ingot was then dumped from the crucible, 

and the sample protrusions that had been cast at the base of the ingot were sheared off and 

collected in aluminum sample containers. The capped sample containers in the hexagonal 

aluminum block were transferred to the Junior Caves (hot cells) in the L&O Building for 

analysis (Stevenson, 1987, PDF pp. 95-129). 

The FCF’s average throughput for the melt-refining process was about 100 kg of fuel per month, 

with a peak throughput of 245 kg/month occurring in mid-1967 (Stevenson, 1987, PDF p. 115). 

Fission products can be divided into three major groups, according to their behavior during melt 

refining. Group 1 includes elements that are either gases or are highly volatile at 1400°C. These 

elements separate from the melt because they are much more volatile than the molten metal 

alloy. Group 2 consists of the elements that react chemically with the ZrO2 crucible at process 

temperature more readily than does uranium. This interaction results in diffusion of the Group 2 

fission products into the crucible matrix, its adherence to the crucible surface, or its separation to 

form a hard slag at the upper-melt surface. Group 3 fission products include those that do not 

separate from the molten alloy because they are not volatile and are less chemically-reactive 

toward ZrO2 than uranium. Fortunately, the Group 3 fission products provide some stability to 

the uranium by reducing swelling during irradiation. The bullet list below shows how the fission 

product elements fall into each of these three groups during melt refining:

• Group 1 (gases or highly-volatile): Krypton, Xenon, Bromine, Iodine, Rubidium, Cesium, and 

Cadmium 

• Group 2 (chemically-reactive with the crucible): Strontium, Yttrium, Tellurium, Barium, 

Lanthanum, Cerium, Praseodymium, Neodymium, Promethium, Samarium, and Europium 

• Group 3 (not volatile and less chemically-reactive): Molybdenum, Ruthenium, Rhodium, 

Palladium, Zirconium, Niobium, Technetium, Silver, Indium, Tin, and Antimony 

Figure 5-35 depicts where all of the key elements go during the melt-refining process. 
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Source: Stevenson, 1987, PDF p. 107

Figure 5-35: Approximate Distribution of Elements in Melt-Refining Process

Hot-Line Step 5 – Skull Recovery: After melt refining, the ZrO2 crucibles, which had a 

uranium-bearing melt residue (a.k.a. skull) adhered to them, were sent to the skull recovery area. 

In the Hot-Line, the skulls were recovered from the ZrO2 crucibles using an oxidation process. 

The ZrO2 crucibles were charged to the skull oxidation furnace. The furnace oxidized the 

enriched-uranium skull adhering to the crucible’s inner wall into a free-flowing powder. The 

oxide was then dumped into a suitable storage can, and the crucible residue discarded as waste. 

Cans of the skull oxide were periodically shipped to the Idaho Chemical Processing Plant (ICPP) 

via a heavily-shielded “oxide coffin” to recover the enriched uranium. The skull oxidation 

furnace was also used to recover uranium from unpourable melt-refining charges and slag 

materials (Hesson, 1963, PDF p. 77; Stevenson, 1987, PDF pp. 98, 102-104) 
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Hot-Line Step 6 – Pin Fabrication: Fuel-pin fabrication involved an injection-casting process. A 

ThO2 (thoria)-coated graphite crucible was charged with a melt-refining ingot, which averaged 

10 kg. Broken pieces of heel from previous injection-casting runs were usually added to the 

charge to reach the proper charge weight, which varied between 11.5 and 14 kg. The charged 

graphite crucible was then placed into the injection-casting furnace. Also located in the furnace, 

in a vertical attitude above the crucible, was a cluster of approximately 100 precision-bored 

thoria-coated Vycor® glass (high-silica glass) fuel-pin molds (a.k.a. Vycor® glass pin molds), 

that were held in position by a pallet stand. Once the casting temperature was attained and the 

charge was molten, the graphite crucible was raised to a position that immersed the lower end of 

the Vycor® glass pin mold cluster. The furnace was then rapidly pressurized to 1.7 atmospheres 

to drive the melt upward into the evacuated molds. The pregnant Vycor® glass pin molds were 

then placed in a feeder bucket for transfer to the pin-processing operation, or for temporary 

storage. The injection-casting heels were removed from the crucibles and used in future castings 

(Stevenson, 1987, PDF pp. 140-141; Fuel Cycle Facility, undated, PDF p. 14) 

A significant step in this process (from a radiological perspective) took place in Room 25 of the 

FCF (the Mold Preparation Room), which was outside the Air and Argon Cells. Because the 

crucibles used in the injection-casting process were graphite, and because graphite is soluble in 

molten uranium, the inside of the graphite crucibles had to be protected with an insoluble coating 

of refractory material to prevent carbon pick-up by the charge. The refractory coating on the 

inside of the graphite crucibles also facilitated skull removal and crucible clean-up. This coating 

was also applied to the outside of the graphite crucibles to reduce radiant-heat losses during the 

injection-casting process. The precision-bored Vycor® glass fuel pin molds also needed to be 

coated with a refractory material to prevent metal adherence and to help create a smooth fuel-pin 

surface. 

From the start of the FCF’s Hot-Line injection-casting process to the start-up of the Cold-Line 

injection-casting process in late-1967, thoria was used as the primary refractory material for 

coating the graphite crucibles and Vycor® glass pin molds (Progress Report, Aug1967, PDF p. 

18; Hesson, 1963; Stevenson, 1987). Only about 35 g of thoria were used per graphite crucible, 

and only about 60 g of thoria were used per 120 Vycor molds (Hesson, 1963, PDF pp. 46, 81, 

160-161, 183; Stevenson, 1987, PDF pp. 134-135, 139-140). By December 1967, zirconia 

(ZrO2) was satisfactorily substituted for thoria in coating graphite crucibles and Vycor® glass pin 

molds (Progress Report, Dec1967, PDF p. 102; Progress Report, Aug1967, PDF p. 18; 

Stevenson, 1987, PDF pp. 139-140). This change eliminated the radiological control problems 

associated with naturally-radioactive thoria. Once Room 25 was decontaminated after this 

change, the potential for thorium exposures at the FCF would have been eliminated. 

Hot-Line Step 7 – Pin Processing: The first step in the pin-processing cycle consisted of 

breaking the pregnant Vycor® glass pin molds away from the cast fuel pins. The molds were 

placed over a slit that was narrower than the molds but wider than the cast fuel pins. A 

pneumatically-driven flat-edge blade forced the mold against the slit, cracking the glass, and 

forcing the fuel pins through the slit. The fuel pins were then sheared to the required length. At 

the length-and-weight measuring station, the pins were checked for weight, length, diameter, 

porosity, and surface defects and voids. (Fuel Cycle Facility, undated, PDF pp. 15-16) 
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Hot-Line Step 8 – Cladding Fabrication and Sodium Loading: Cladding fabrication and sodium 

loading were two non-radiological processes that were performed outside of the FCF’s Air and 

Argon Cells. The completed jacket assemblies (empty fuel-cladding tubes) were visually 

inspected for evidence of superficial damage (e.g., faulty welds, scratches) and leak-checked. 

Acceptable jacket assemblies were then loaded with sodium in an argon-filled glovebox located 

in FCF Room 23. Batches of sodium-loaded jackets were sealed in polyethylene bags before 

transferring them from the glovebox in Room 23 to the Argon Cell (Fuel Cycle Facility, undated, 

PDF pp. 16–17). 

Hot-Line Step 9 – Fuel Element Assembly: All final fuel-element-closure operations were 

conducted remotely in the Argon Cell. After the sodium-loaded jackets were received in the 

Argon Cell, fuel pins were then inserted in their respective jackets and the sodium was melted in 

a shell furnace. End plugs (caps) were inserted into the jackets and peripherally-welded to the 

jackets. The finished product (now called a fuel element) was visually inspected for welding 

flaws. With the fuel material and bond sodium protected by the end plugs, the finished elements 

were transferred to the Air Cell for leak-testing (Fuel Cycle Facility, undated, PDF pp. 17). 

Hot-Line Step 10 – Leak Testing: The closure welds on the newly-welded fuel elements were 

leak-tested for integrity with the helium-pressure leak detector located in the Air Cell. The 

accepted elements were placed in the bonding magazine located in the bonder magazine 

turntable. (Stevenson, 1987, PDF pp. 177–178) 

Hot-Line Step 11 – Sodium Bonding: All finished elements were subjected to an impact-bonding 

process. Following this process, the elements were cooled in a manner that allowed the bond 

sodium to freeze from the bottom upward, thus eliminating the possibility of shrinkage voids and 

contractional tearing (Fuel Cycle Facility, undated, PDF p. 18). 

Hot-Line Step 12 – Sodium Bond Testing: The final step in testing activities consisted of an 

examination of bond quality. Elements, one at a time, were raised upward through an 

eddy-current coil that sensed voids as an electrical disturbance in an otherwise passive signal. 

The acceptable elements were then removed from the magazine and placed in a 91-element 

storage holder. When the 91-element holder was filled, it was transferred to the final assembly 

station for loading into a subassembly. Elements rejected for one reason or another were fed 

back to the fuel stream by stripping away the jackets and returning the fuel material to the front 

end of the melt-refining cycle. (Stevenson, 1987, PDF pp. 178–179; Fuel Cycle Facility, 

undated, PDF pp. 19-20) 

Hot-Line Step 13 – Subassembly Fabrication: In the Air Cell, completed fuel elements were 

loaded into fuel assemblies using master-slave manipulators. The completed subassemblies 

were tensile-tested and tested for straightness prior to going to the EBR-II Reactor (Fuel Cycle 

Facility, undated, PDF p. 20). 

Hot-Line Step 14 – Scrap Reclamation & Waste Disposal: Two scrap streams from the process 

contained a sufficient quantity of uranium to warrant recovery. The first stream was the mixture 

of uranium and fission product oxides obtained from the used melt-refining crucibles, and 

uranium oxide in slags from scrap consolidation or alloy-preparation melts. The uranium from 

this stream represented from 3% to 8% of the total uranium in the charge (Stevenson, 1969). As 
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previously indicated in Step 5, cans of the uranium oxide were periodically shipped to the ICPP 

via a heavily-shielded “oxide coffin” to recover the enriched uranium (Hesson, 1963, PDF p. 77; 

Stevenson, 1987, PDF pp. 98, 102-104). A large part of the fission-product activity is assumed 

to have gone with the skull oxide sent to the ICPP (Stevenson, 1987, PDF p. 212). The second 

stream consisted of crushed Vycor® glass pin molds containing on the order of 1% of the 

uranium charged to injection-casting as uranium-fissium alloy. To facilitate concentration of 

FCF effort on fuel production, the crushed Vycor® glass pin molds were also shipped to the ICPP 

for uranium recovery by acid leaching (Stevenson, 1969).

Process wastes, except for the caustic solution obtained by washing sodium-contaminated 

subassemblies, were all in form of solids or gases. The principal solid radioactive wastes were:

• Activated stainless-steel components of the fuel subassemblies that were removed in 

disassembly

• Activated stainless-steel cladding (and possibly some chips of fuel) and adherent sodium that 

was separated from the fuel pin in the de-cladding operation 

• Zirconia crucibles, fume traps, and graphite ingot molds used to melt and cast the ingot of 

purified fuel in melt refining

• Graphite crucibles and stainless-steel mold pallets used to re-melt the purified ingot and to 

cast fuel rods by injection-casting

• Various contaminated tools and defective in-cell equipment

All solids, including the stainless-steel subassembly hardware, are placed in 11-inch diameter by 

6-foot high steel containers equipped with lug clamp lids. Some of the wastes are secondarily 

contained in smaller steel containers with similar lids. These containers were then sent to the 

Radioactive Scrap and Waste Facility (RSWF) for disposal. (Stevenson, 1987, PDF p. 212; 

Stevenson, 1969) 

The principal radioactive gaseous wastes were the off-gases from the melt refining, skull 

oxidation, and injection-casting furnaces; the scrubbing gas stream from sodium removal; the 

contaminated argon exhausted from transfer locks; the air cell exhaust; and the potentiallycontaminated building exhaust from operating areas. Figure 5-36 depicts the gaseous effluent 

treatment system for the FCF’s Air and Argon Cells. Iodine is largely absorbed, and krypton and 

xenon are held for up to 30-days’ decay before release through a 200-foot stack under favorable 

wind conditions. Any residual iodine is absorbed on a hot AgNO3 bed (a.k.a. silver tower). 

(Stevenson, 1987, PDF p. 212; Stevenson, 1969) 
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Source: Hesson, 1963

Figure 5-36: Gaseous Effluent Treatment System for FCF Air and Argon Cells

Overview of the FCF Cold-Line Fuel Production Process 

Cold-Line (unirradiated) fuel-production processes (which took place outside the FCF’s Air and 

Argon Cells) were very similar to the Hot-Line processes; the only major differences being in the first 

four steps. The Cold-Line process did not include the Transfer of Subassemblies to FCF, Sodium 

Removal, Disassembly, Decanning, and Pin Chopping steps. The Cold-Line’s alloy preparation was 

very similar to the Hot-Line’s melt-refining step, with the main difference being that fission products 

and other impurities were not in the Cold-Line’s feedstock (Stoddart, 1967; Hass, 1979). Therefore, 

Cold-Line fuel production only involved 12 major process steps. 

The 12 Cold-Line fuel production process steps are summarized below. 

NOTE: The Hass document (Hass, 1979) appears to only describe Cold-Line fuel production at 

ANL-E versus the Cold-Line fuel production at ANL-W. Even though the processes at 

ANL-E and ANL-W were nearly identical, there were some differences. The most 

notable differences included: (1) the mold coatings; (2) alloy preparation crucibles; and 
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(3) the make and design of some of the equipment. It should also be noted that many of 

the pictures provided in the Hass document appear to be from the Cold-Line facilities at 

ANL-E versus ANL-W. 

Cold-Line Step 1 – Alloy Preparation: For the Cold-Line melt runs, either 10 kg or 20 kg charges 

were loaded into a new zirconia (ZrO2) crucible. Each charge was melted to create the alloy, and 

then tilt-poured into a graphite ingot mold. After cooling, the ingots were removed, weighed, 

sampled, and held for injection-casting of the fuel pins. (Stoddart, 1967; Hass, 1979) 

Prior to each run, the graphite ingot mold was coated with either a thoria/ethyl alcohol or 

zirconia/ethyl alcohol wash to preclude the melt from reacting with and degrading the graphite 

ingot mold. The Reactor Development Program Progress Reports for September and December 

1967 indicate that thoria was being used through September 1967; however, by December 1967, 

the use of thoria had been phased out (Progress Report, Dec1967, PDF p. 102; Progress Report,

Sept1967, PDF p. 19). 

Cold-Line Step 2 – Skull Recovery: After alloy preparation, the crucibles, which had a uraniumbearing melt residue (a.k.a. skull) adhered to them, were processed to remove the skulls. Early 

Cold-Line production information indicates that a mechanical process was used, and that process 

may have been as simple as breaking the crucibles, or may have involved chipping the skull 

material out with a hammer and screwdriver (Stoddart, 1967). A Health Physics Report for 

October 1967 indicates that this process was performed in FCF Room 26 (Health Physics 

Monthly Report, Dec1968, PDF p. 23; Health Physics Monthly Report, Jan1968, PDF p. 25). 

Cold-Line Step 3 – Pin Fabrication: The Cold-Line’s pin-fabrication process, which was 

performed in FCF Room 20, was nearly identical to the Hot-Line’s pin-fabrication process that 

was performed in the Argon Cell (Hass, 1979, PDF pp. 15-20). Details of this process were 

previously described in Hot-Line Step 6.

Cold-Line Step 4 – Pin Processing: The Cold-Line’s pin-processing step was also performed in 

FCF Room 20 (Health Physics Monthly Report, Dec1968, PDF p. 14). After the cast fuel pins 

had cooled, they were moved to a hood, where the Vycor® glass pin molds were broken and the 

fuel pins removed. The pins were sheared to length in preparation for inspection. The pins were 

then checked for length, weight, diameter, internal flaws, texture, and compositional 

requirements. (Hass, 1979, PDF pp. 20-22) 

Cold-Line Step 5 – Cladding Fabrication and Sodium Loading: The Cold-Line’s claddingfabrication and sodium-loading processes were identical to Hot-Line’s processes, as previously 

described in Hot-Line Step 8, with the only exception being that materials were not transferred to 

the Argon Cell (Hass, 1979, PDF pp. 22-31). As with the Hot-Line, all sodium-loading 

operations were carried out in an argon-filled glovebox located in FCF Room 23. 

Cold-Line Step 6 – Fuel Element Assembly: The Cold-Line’s fuel element assembly was 

performed in a glovebox with an inert-argon atmosphere, instead of the FCF’s Argon Cell (Hass, 

1979, PDF pp. 27-34). It is uncertain where this process step was performed. However, based 

on the available survey records, this step was most likely performed in the Cold-Line glovebox 

located in FCF Room 23 (Survey Reports, Feb1969). The general process was similar to the 

Hot-Line’s processes, as previously described in Hot-Line Step 9 (Hass, 1979, PDF pp. 35–35). 
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Cold-Line Step 7 – Leak Testing: Because the newly-welded fuel elements would need to be 

leak-tested before they were removed from the inert-argon atmosphere, the leak-testing was 

likely also performed in the Cold-Line glovebox located in FCF Room 23. The general process 

was similar to the Hot-Line’s processes, as previously described in Hot-Line Step 10 (Hass, 

1979, PDF pp. 36–38). 

Cold-Line Step 8 – Sodium Bonding: Based on the available survey records, this step was 

performed in the Low Bay of the Inspection Test Facility (Survey Reports, Feb1969). The 

general process was similar to the Hot-Line’s processes, as previously described in Hot-Line 

Step 11 (Hass, 1979, PDF pp. 38–39). However, the process for producing Mark-II fuel 

elements included the extra step of dimpling the fuel elements. Because the Mark-II fuel 

elements did not rely on internal restrainers to keep the fuel pins in position within the elements, 

three dimples were impressed into the Mark-II fuel elements 120o circumferentially apart from 

each other (Hass, 1979, PDF p. 40). 

Cold-Line Step 9 – Sodium Bond Testing: Based on the available survey records, this step was 

performed in the Low Bay of the Inspection Test Facility (Survey Reports, Feb1969). The 

general process was similar to the Hot-Line’s processes, as previously described in Hot-Line 

Step 12 (Hass, 1979, PDF pp. 41–39). 

Cold-Line Step 10 – Element Final Inspection: This was an added step that was not performed in 

the Hot-Line. Based on the available survey records, this step was performed in a new room that 

was added on to the southwest end of the Inspection Test Facility (Survey Reports, Feb1969). 

All fuel elements were inspected for dimensional conformance and surface quality. Special 

calibrated inspection gauges were used to verify conformance-to-length, straightness, and dimple 

location. All fuel element cladding surfaces had to be free of splits, cracks, seams, and 

inclusions and also had to be free of scratches, dents, or off-specification marks (Hass, 1979, 

PDF pp. 46–47). In addition, one fuel element was selected from each batch for destructive 

evaluation. The cladding from that element was stripped from the fuel and prepared for 

metallographic examination. The bare fuel pins were evaluated for texture, grain size, and 

extraneous phases (Hass, 1979, PDF p. 47). 

Cold-Line Step 11 – Subassembly Fabrication: This step was performed in the High Bay of the 

Inspection Test Facility. The general process was assumed to have been similar to the HotLine’s processes, as previously described in Hot-Line Step 13. 

Cold-Line Step 12 – Scrap Reclamation & Waste Disposal: Because unirradiated feedstock was 

used for the Cold-Line, a much higher percentage of the Cold-Line materials could be reclaimed 

and recycled than from the Hot-Line. In the Cold-Line, the primary uranium reclamation was 

from the recovery of, and reuse of, the alloy-preparation skulls and injection-casting heels, which 

were in forms that could be returned directly back into the Cold-Line processes. No information 

is available on what was done with the crushed Vycor® glass pin molds from the Cold-Line, 

which would have contained about the same amount of residual uranium as the crushed glass 

from the Hot-Line. Because no acid-leaching process was set up at ANL-W to recover that 

uranium, it is assumed that crushed Vycor® glass pin molds from the Cold-Line were also 

shipped to the ICPP for uranium recovery by acid-leaching. 
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Details regarding waste disposal for the Cold-Line are not available.

FCF Timeline Through 1979 

The timeline in Table 5-3 summarizes important dates in the FCF’s operating history through 1979. 

Table 5-3: Important Dates in FCF Operating History

Date Activity

December 1962 Construction of the FCF completed (Stevenson, 1987, PDF p. 15). 

March 1963 Air Cell significantly contaminated from removal of a several-thousand-curie antimony 

source for the EBR-II Reactor from its capsule, and while attaching it to a suitable fixture 

for handling it in the reactor. Contamination was due to some loose highly active deposit 

on the surface of the tantalum-clad source. Decontamination of the Air Cell was performed 

and completed. (Progress Report, Mar1963, PDF p. 30) 

August 1963 Start of operations in the Argon Cell (Research Highlights, 1965, PDF p. 22). The first 

melting of uranium (depleted) was accomplished satisfactorily in the melt-refining furnace 

system (Progress Report, Aug1963, PDF p. 18).

Early-1964 Installation of process equipment was completed. Unirradiated dummy fuel elements and 

subassemblies containing unenriched uranium fuel alloy were fabricated and carried 

through all the process steps to determine where modifications might be needed and to 

verify equipment operability. (Stevenson, 1987, PDF p. 248)

January 31 to February 

27, 1964

On January 31, 1964, the Argon Cell was filled with air to allow personnel to enter the cell 

to perform maintenance on cranes and manipulators, to speed installation of additional 

equipment, and to remove mercury which had spilled into the cell from a blown out blister 

seal located near the cell’s ceiling. On February 27th, the Argon Cell was refilled with 

argon. (Progress Report, Feb1964; Progress Report, Mar1964; Research Highlights, 1965, 

PDF p. 23)

March 1964 All the fuel element fabricating equipment in the Argon Cell became operational (i.e. the 

melt-refining equipment, injection-casting equipment, pin-processing equipment, and fuel 

element settling and welding equipment). Fuel element can now be made in the Argon Cell. 

(Progress Report, Mar1964) 

September 1964 One core subassembly and one blanket subassembly were transferred from the EBR-II 

Reactor to the FCF, and processing of the first irradiated reactor fuel from the EBR-II 

reactor was initiated in the FCF’s Hot-Line (Progress Report, Sep1964, PDF p. 52; 

Stevenson, 1987; Koch, undated; 148941). 

April 1965 The manufacturing of the first subassembly containing irradiated fuel was completed. 

(Koch, undated; Stevenson, 1987, PDF p. 248)

June 1965 The first two cans of FCF waste were shipped to the Radioactive Scrap and Waste Facility 

(RSWF), which was also the first use of the RSWF (Progress Report, Jun1965, PDF p. 12). 

April 1967 To reduce the number of transfers through the Air Cell, a new opening was made in the wall 

of the transfer subcell between the Air and Argon Cells. The creation of the new openings 

included the installation of a window and transfer lock. This transfer system was created to 

allow all the unirradiated materials, such as crucibles, molds, and waste cans, to be 

introduced directly into the Argon Cell via the large lock rather than having to traverse the 

whole length of the Air Cell. When the mobile decontamination facility is completed, this 

lock system will also be used to remove equipment components for cleaning and repair. 

(Progress Report, Apr1967, PDF p. 30) 
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Date Activity

July 1967 Installation of the FCF Cold-Line’s alloy-preparation furnace was completed. Five test runs 

were completed using depleted uranium-fissium alloy. (Progress Report, Jul1967, PDF p. 

66)

Because the FCF’s in-cell storage capacity is now fully utilized, subassemblies being 

fabricated with unirradiated fuel are being stored in the L&O Building vaults (Progress 

Report, Jul1967, PDF p. 65). 

November 1967 The Cold-Line equipment became available for routine operations, and routine Cold-Line 

production began (Progress Report, Nov1967, PDF pp. 108-109).

January 1968 The first four uranium-5 w/o fissium ingots manufactured by a commercial vendor were 

received by ANL. The ingots were used in the FCF’s Cold-Line injection-casting furnace. 

(Progress Report, Jan1968, PDF p. 96) 

April 1968 Construction of the truck lock and the second-story addition to the FCF office building was 

started. The truck lock provided a ventilation barrier for loading and unloading casks from 

trucks, and an area for decontaminating truck beds and casks. (Progress Report, Apr1968, 

PDF p. 104)

The first neutron radiographs were performed in the Argon Cell (Progress Report, Apr1968

PDF p. 92).

, 

October 1968 Announcement of programmatic changes for the FCF, and that there will be a gradual 

decrease in Hot-Line fuel production and an increase in support on the experimental 

irradiation programs (Progress Report, Oct1968, PDF p. 94). 

December 1968 Modifications of the transfer cell (a.k.a. Subcell 4) below the Air and Argon Cells were 

completed. These included: (1) placing into operation the pneumatic transfer system for 

conveying samples to the analytical caves in the L&O Building; (2) installing and testing 

the movable shield plug for the entrance to the subcell; (3) installing two master-slave -

manipulators; and (4) installing additional lighting equipment in the subcell. (Progress 

Report, Dec1968, PDF pp. 75, 84) 

1969 Permanent decontamination and repair facilities were constructed on the FCF roof and in 

the FCF basement (Cook, 1969, PDF p. 17).

March 1969 Phasing out of Hot-Line production activities in the Argon Cell was completed (Progress 

Report, Mar1969). Fuel elements remaining in the Air Cell were being processed for use in 

subassemblies (Progress Report, Mar1969). During the operation of the FCF’s Hot-Line

roughly 35,000 fuel elements (the equivalent of 5 EBR-II Reactor core loadings) were 

recycled and returned to the EBR-II reactor, in the form of over 400 fuel subassemblies

(Cook, 1975; Stevenson, 1987, PDF p. 260). 

Cold-line equipment was modified for fabrication and inspection of Mark-II fuel elements. 

Preparation of uranium-fissium metal ingots for the Mark-II fuel alloy started in the alloypreparation furnace (FCF Room 26) of the Cold-Line. (Progress Report, Mar1969)

December 1969 Cold-Line fuel pin and fuel element production equipment put on stand-by (Progress 

Report, Dec1969, PDF p. 87). However, Cold-Line continued to fabricate unirradiated 

subassemblies from vendor produced fuel elements and perform impact-bonding of the 

unbonded vendor fuel elements (Progress Report, Dec1969, PDF p. 87). Fuel-pin 

fabrication in the FCF was not reinitiated until 1982 (Batte, 1986, PDF p. 9).

1969 The first refurbishment of the Air Cell, which was assumed to take place after all irradiated 

fuels had been removed from the Air and Argon Cells (HFEF History, 1979).

July 1970 Reactor Development Program Progress Report for July 1970 identifies the FCF as being 

part of both the Hot Fuel Examination Facilities and EBR-II Operations (Progress Report, 

Jul1970).
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Date Activity

August 1970 The FCF was renamed as the Fuel Examination Facility (FEF), to reflect the change in 

emphasis and nature of the work being done in the facility and planned for it (Progress 

Report, Aug1970). However, after that date, documents still refer to this facility as being 

both the FCF and FEF, sometimes in the same document.

1972 The FCF was renamed the Hot Fuel Examination Facility-South (HFEF-S) after 

construction on the Hot Fuel Examination Facility-North (HFEF-N) was completed. 

(Stevenson, 1987, PDF p. 260) 

Early-1973 FCF Argon Cell was remotely decontaminated in preparation for RAS-TREAT sodium loop 

experiment work. The RAS-TREAT sodium loop experiments were sponsored by ANL-E’s 

Reactor Analysis and Safety Division, and were irradiated in the TREAT Reactor. (Cook, 

1975, PDF p. 24) 

July 1973–April 1974 Several plutonium bearing Mark-IIA loop experiments were sectioned in the FCF’s Argon 

Cell as part of the posttest support of the RAS-TREAT loop experiments. The purpose of 

sectioning the loops is to reduce the loop experiments into smaller segments for shipment to 

the Materials Science Division (MSD) at ANL-E, where the detailed posttest examination 

of the loop experiments is performed. During this period, loops L3, L4, E6, E7, H4, and H5 

were sectioned and their ALIPs (annular linear-induction pumps) were recovered. This 

activity contaminated the Argon Cell with plutonium. (Cook, 1975)

1976 The FCF equipment mockup/qualification area was substantially enlarged (HFEF History, 

1979).

The second refurbishment of the Air Cell was performed in 1976 (HFEF History, 1979).

1977 Argon Cell was shut down for to correct several problems that had developed over its 13-

year life. Principally, the overhead handling systems had deteriorated, but in-cell 

illumination and the viewing windows also needed improvement. (Stevenson, 1987, PDF p. 

260) 

June 23, 1978 First entry into the Argon Cell since February 1964 (HFEF History, 1979; Entry Log; 1979;

Research Highlights, 1965, PDF p. 23). Operations involving irradiated fuels began in the 

Argon Cell began after February 1964 (Entry Log; 1979). The Argon Cell was entered in 

February 1964, which was after some work involving unirradiated dummy fuel elements 

and subassemblies containing unenriched uranium fuel alloy was performed in the Argon 

Cell (Stevenson, 1987, PDF p. 248; Research Highlights, 1965, PDF p. 23). 

1987 By 1987, operations in the Argon Cell had not restarted. (Stevenson, 1987, PDF p. 260)
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5.1.2.4 Fuel Assembly & Storage Building (FASB) 

ATTRIBUTION: Section 5.1.2.4 was completed by Brian Gleckler, Dade Moeller, Inc. All 

conclusions drawn from the data regarding FASB were peer-reviewed by the individuals listed on the 

cover page. The rationale for all conclusions in subsequent sections of this document devoted to 

FASB are explained in the associated text.

In Figure 5-24, the FASB is Building 787 located adjacent to the north side of the FCF. The FASB is 

also known by the following building numbers: ANL-787, EBR-II-627, and MFC-787. The building 

was also sometimes referred to as the “FAS Building.” In later years (i.e., sometime after 1979), this 

facility was renamed the Fuels and Applied Science Building, which preserved the acronym “FASB.” 

Figure 5-37 is based on a radiological survey map that shows the general FASB layout. 

Construction of this 6,025 ft2 (560 m2) building was completed in 1971 (Site Plan, 1981). The 

structure features masonry exterior walls, built-up roof on pre-stressed concrete beams, a reinforced 

concrete floor, and no windows (Site Plan, 1981). The building was equipped for assembling EBR-II 

driver-fuel pins into fuel elements, fabricating EBR-II driver and experimental subassemblies, 

performing inspections on fuel elements and finished subassemblies, and storing fuel elements and 

finished subassemblies (Site Plan, 1981). The facility was also use to perform air-flow tests on 

subassemblies (Site Plan, 1981). 

Based on radiological survey records, Cold-Line (unirradiated) fuel production at the FASB appears to 

have begun around February 1972. An October 26, 1972 safety review for the FASB’s Fuel Storage 

Vault indicates that the vault was not used until sometime after October 1972 (FASB Vault, 1972). 

Source: FASB Surveys, 1983, derived from figure on PDF p. 35

Figure 5-37: General Layout of FASB
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5.1.2.5 Inspection & Test Facility (ITF)

ATTRIBUTION: Section 5.1.2.5 was completed by Brian Gleckler, Dade Moeller, Inc. All 

conclusions drawn from the data regarding ITF were peer-reviewed by the individuals listed on the 

cover page. The rationale for all conclusions in subsequent sections of this document devoted to ITF 

are explained in the associated text.

In Figure 5-24, the ITF is Building 772 located approximately 350 ft. (107 m) east of the FCF. The 

ITF is also known by the following building numbers: ANL-772, EBR-II-616, and MFC-772. 

Construction of this 3,455 ft2 (320 m2) building was completed in 1966 (Site Plan, 1981). The 

structure features masonry exterior walls, steel framing, and corrugated sheet metal siding on the 

high-bay area (Site Plan, 1981). Figure 5-38 is based on a radiological survey map showing the 

general ITF layout.

Source: ITF Survey Data, 1976, derived from figure on PDF p. 29

Figure 5-38: General Layout of ITF

The ITF was originally constructed to inspect and test EBR-II reactor components (including 

in-sodium tests), and to perform nondestructive testing of experimental irradiation specimens and 

reactor components (Cold-Line Facility, 1969; Site Plan, 1981). By August 1967, those activities 

were curtailed by the pressing demand for a temporary Cold-Line fuel production facility, which 

included parts of the ITF and FCF (Cold-Line Operation, 1967; Cold-Line Facility, 1969; Site Plan, 

1981). Because of that effort, most of the high-bay and half of the low-bay of the ITF were converted 

for Cold-Line fuel production (Cold-Line Facility, 1969). Based on a review of ITF radiological 

survey records, most of its Cold-Line work involved installing fuel elements and experiments into 

subassemblies for the EBR-II Reactor. The radiological survey records also indicate that irradiated 

fuels and materials were sometimes handled at the ITF. 
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By July 1969, plans were being made to convert the ITF back to its original tasks as soon as possible 

(Cold-Line Facility, 1969). Cold-Line operations at the ITF appear to have been significantly scaled 

back after the permanent Cold-Line fuel production facility (FASB) started operating in February 

1972. After fuel production operations began at FASB, radiological survey records and health physics 

log books indicate that the Cold-Line operations at the ITF were reduced to performing X-ray 

radiography on fuel pins. 

Prior to the HFEF-N’s NRAD Facility becoming operational in March 1978, the 20-ft (6.1-m) deep pit 

in the ITF’s high-bay area was fitted with a tank for receiving the NRAD Reactor fuel (Neutron 

Radiography, 1979; Site Plan, 1981). Radiological survey records indicate that the pit in the high-bay 

may have been present when the ITF was constructed, and initially may have been used to perform 

X-ray radiography on subassemblies (ITF Survey Data, PDF pp. 20-21). 

By February 1981, ITF activities included: (1) testing non-radioactive EBR-II components using 

X-ray, ultrasonic, eddy-current, thermal, and other nondestructive methods; and (2) rigging, tackle, 

and load testing (Site Plan, 1981). 

5.1.2.6 Hot Fuel Examination Facility-North (HFEF-N)

ATTRIBUTION: Section 5.1.2.6 was completed by Brian Gleckler, Dade Moeller, Inc. All 

conclusions drawn from the data regarding HFEF-N were peer-reviewed by the individuals listed on 

the cover page. The rationale for all conclusions in subsequent sections of this document devoted to 

HFEF-N are explained in the associated text.

HFEF-N is a large hot-cell facility consisting of heavily-shielded cells, support laboratories, and 

support equipment (HFEF Training, 2009; HFEF Design, 1976). The facility was specifically 

designed to remotely characterize highly-irradiated fuel and structural materials (HFEF Training, 

2009). Radiological operations started at HFEF-N in August 1972 and continue to the present. 

The initial function of the HFEF-N was to prepare and examine irradiation experiments in support of 

the Liquid Metal Fast Breeder Reactor (LMFBR) Program (HFEF Safety Report, 1975). To support 

LMFBR fuel and structural material development, the materials examined at the HFEF-N were 

usually irradiated in the EBR-II Reactor (HFEF Safety Report, 1975). To support the LMFBR safety 

programs, HFEF-N was used to prepare test specimens and examine sodium loop experiments 

irradiated in the TREAT Reactor, INL’s Engineering Test Reactor (ETR), or INL’s Power Burst 

Facility (PBF) (HFEF Safety Report, 1975). Also, at times, irradiated materials from other programs 

and facilities were accommodated at HFEF-N (HFEF Safety Report, 1975). 

In Figure 5-24, HFEF-N is Building 785, and is located approximately 175 ft. (53 m) north of the 

FCF. The HFEF-N is also known by the following building numbers: ANL-785, EBR-II-626, and 

MFC-785. After HFEF-S was renamed the Fuel Conditioning Facility in 1993, HFEF-N was only 

referred to as HFEF, which was what the FCF was sometimes referred to prior to the HFEF-N being 

constructed. Figure 5-39 shows an elevation cross-section of HFEF-N. Figure 5-40 is a floor plan for 

the HFEF-N Operating Floor. Figure 5-41 is an elevation detail of the NRAD Reactor and 

Radiography Room at HFEF-N. 
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Source: HFEF Design, 1976, derived from figure on PDF p. 20

Figure 5-39: Elevation Cross-Sections of HFEF-N 
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Source: HFEF Design, 1976, derived from figure on PDF p. 21

Figure 5-40: Floor Plan of HFEF-N Operating Floor
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Source: Bess, 2014, PDF p. 16

Figure 5-41: Elevation Detail of the NRAD Reactor and Radiography Room at HFEF-N 

Facility Description

HFEF-N is structurally designed to accommodate support equipment on all sides of the centrallylocated hot cells. This large heavily-shielded hot-cell facility includes two adjacent hot cells (the 

Main Cell and Decon Cell). The building is framed with concrete that is heavily-reinforced with steel 

beams. From ground-level to the top of the second-floor area, the exterior walls are concrete block. 

From the top of the second floor upward, the exterior walls are covered in sheet metal, insulation, and 

an interior lining. The building floors are made of steel-reinforced concrete (HFEF Design, 1976). 
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The three-story HFEF-N building is 112 ft. wide x 154 ft. long, has a gross floor area of 56,570 ft2, 

and a gross volume of 1,337,200 ft3. The building is designed with ample space and heavy-duty 

handling equipment to accommodate large irradiated loops, such as the Sodium Loop Safety Facility 

(SLSF), loops from INL’s Engineering Test Reactor, as well as smaller equipment for handling 

reactor subassemblies (see Figure 5-39). Numerous penetrations into the cells provide easy access in 

transporting and handling irradiated materials into and out of the heavily-shielded cells. In-cell 

equipment and out-of-cell laboratories provide for both destructive and nondestructive examination of 

the irradiated reactor components. (HFEF Design, 1976) 

HFEF-N was designed as a post-irradiation examination facility and is fully equipped for destructive 

and nondestructive examinations. Following receipt of irradiated test materials or components, the 

specimens can be visually inspected, measured, documented, and subjected to gamma scanning, 

ultrasonic testing, neutron radiography, and other examinations. Destructive testing begins with 

cutting and preparation for metallographic and ceramographic optical examination, micro-hardness 

testing, and mechanical properties testing. Further preparation can be performed and samples 

transferred to other facilities for continued analytic chemistry or electro-optical analysis using transfer 

casks or a pneumatic transfer system between facilities (HFEF Training, 2009). 

As shown in Figure 5-39, the HFEF-N building has four floors: (1) Service Floor (basement); (2) 

Operating Floor; (3) Second Floor; and (4) High-Bay Floor (HFEF Design, 1976): 

The major HFEF-N radiological work areas are listed below and described in the following 

subsections. 

• Main Cell

• Decon Cell

• Out-of-Cell Metallographic Areas

• Radiation Safety Laboratory 

• Third Floor Repair Area 

• NRAD Facility

• Truck Lock

• Cask Tunnel 

• Transfer Tunnel 

Main Cell

The heart of the HFEF-N facility where most of the examinations were carried out is the Main Cell 

(HFEF Safety Report, 1975). The Main Cel1 provides facilities for disassembling, examining, and 

testing irradiated materials in an inert (argon gas) atmosphere (HFEF Design, 1976). In anticipation 

of handling plutonium-bearing experiments, the Main Cell was designed for the particle-tight 

containment of any plutonium contamination (Site Plan, 1981). 

The Main Cell has 15 workstations that are each 10-ft wide, and numbered “1M” through “15M.” 

Each workstation can be equipped with a pair of manipulators, a shielding window, lights, a utility 

distribution system, examination equipment, and/or suitable work tables. A few work stations are 

equipped with ports for specific purposes (e.g., periscopes, stereomacroscopes, and gamma-scanning 

equipment). Work stations that are not provided with examination equipment at start-up can be fitted 
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with sealed and plugged ports for utilities. These items can be added later without releasing 

radioactive contaminants, polluting the cell atmosphere, or extensively interrupting work at adjacent 

stations (HFEF Safety Report, 1975). 

Decon Cell

The Decon Cell is a decontamination hot cell with an air atmosphere that is equipped for 

decontaminating and packaging nuclear fuel capsules and elements for shipment (HFEF Design, 

1976). It also provides a shielded path for transfers of radioactive material to and from the Main 

Cell’s inert atmosphere, and has work stations for some sodium-loop operations (HFEF Design, 

1976). The Decon Cell is separated from the west end of the Main Cell by a 4-ft-thick concrete wall 

(HFEF Safety Report, 1975; HFEF Design, 1976). The cell has outer-shielding walls equivalent to 

those of the Main Cell and has inside dimensions of 30 ft. wide x 20 ft. long x 25 ft. high. The cell 

floor is lined with stainless steel and the walls are lined with carbon steel to a height of 13.5 ft. above 

the cell floor. Wall, roof, and floor penetrations are provided for windows, utility-service 

feedthroughs, special feedthroughs, in-cell handling equipment, ducting, and transfer hatches. These 

penetrations are steel-lined and surrounded by high-density shielding inserts (HFEF Design, 1976). 

The Decon Cell has six work stations that are numbered “1D” through “6D.” Some are used for 

decontamination of equipment, irradiated components, and loop hardware; others are used for 

transfers and future projects (HFEF Safety Report, 1975). 

The Decon Cell is maintained at a negative pressure relative to the operating corridor, the cask tunnel, 

the high-bay area, and the transfer tunnel to prevent the release of particulate contamination from the 

cell. Remotely-replaceable filters and de-misters are installed in the exhaust ducts within the cell. 

The spread of vapors from decontamination reagents is minimized by performing wet operations 

within closed and vented vessels such as the spray chamber (HFEF Safety Report, 1975). 

Out-of-Cell Metallographic Areas

The Out-of-Cell Metallographic Areas are used for all but the high-level metallographic specimens. 

These areas are on the Operating Floor across the north operating corridor from the Main Cell (see 

Figure 5-36). These Metallographic Areas consist of four sub-areas (HFEF Design, 1976):

1. The Metallograph Room (Room 123) houses the metallograph loading cell and the scanning 

electron microscope (HFEF Design, 1976). 

2. The Low-Level Metallograph Lab is a complex of rooms used for preparation of low-level, 

metallographic samples and unirradiated fuels. It includes the metallograph lab (Room 125), 

cutting room (Room 127), and macrophoto room (Room 129) (HFEF Design, 1976).

3. The Clean Metallograph Lab (Room 128) is used for preparation and microscopic examination of 

cold non-fissile metallographic specimens (HFEF Design, 1976). 

4. The Photo Lab (Rooms 121, 122, 124, and 126) provides the support necessary for photographic 

examinations conducted during hot-cell and out-of-cell lab operations (HFEF Design, 1976). 
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High-level or “hot” metallographic specimens (fueled and non-fueled capsules and elements and other 

irradiated components) are prepared remotely in the in-cell containment box located at station 2M of 

the Main Cell (HFEF Safety Report, 1975). 

Metallographic samples are prepared by means of cutting, mounting, grinding, polishing, and etching 

operations. Then both low-level and high-level specimens are examined by use of a remotelyoperated, shielded optical metallograph. This gastight metallograph, with its gastight shielded 

enclosure (the metallograph loading box) is located in the Metallograph Room. Sample transfers 

between the in-cell containment box and the metallograph loading box are accomplished by use of a 

gastight and shielded pneumatic transfer system (HFEF Safety Report, 1975). 

Radiation Safety Laboratory 

The Radiation Safety Laboratory (Room 130) is on the Operating Floor across the north operating 

corridor from the Decon Cell (see Figure 5-36). This lab is used to detect and contain the spread of 

contamination into the environment or prevent undue radiation exposure to personnel (HFEF Design, 

1976). 

Hot Repair Area

The Hot Repair Area (a.k.a. Third Floor Repair Area, Roof Repair Area, and/or Repair Area) is a 

complex of rooms on the High-Bay Floor (third floor) directly above Decon Cell and the west end of 

the Main Cell (see Figures 5-35 and 5-37). The outside dimensions of this facility are 45 ft. x 46 ft.

(HFEF Design, 1976). The Hot Repair Area includes two primary work rooms, the Hot Repair Room 

and Suspect Repair Room. It also includes the following work and support rooms: glovebox room, 

equipment access and transfer area, isolation area, waste packaging area, service room, and change 

room (HFEF Safety Report, 1975; HFEF Design, 1976).

The Hot Repair Area provides a containment enclosure in which radioactively-contaminated 

equipment or suspect (potentially-contaminated) equipment can be repaired or modified by personnel 

wearing protective clothing (HFEF Safety Report, 1975; HFEF Design, 1976). These operations 

include rebuilding of mechanical master-slave manipulators, repair of Main Cell and Decon Cell 

telescoping-tube and crane carriages, and maintenance or modification of examination equipment 

(HFEF Safety Report, 1975). 

NRAD Facility

Through 1979, the Neutron Radiography Facility (NRAD or N-RAD) was mostly located on the 

HFEF-N Service Floor below the Main-Cell (see Figures 5-38 and 5-39). As shown in Figure 5-36, 

the NRAD Reactor control room is located in Room 112 on the Operating Floor (HFEF Design, 

1976). NRAD uses a 250-kW TRIGA® (Training, Research, Isotopes, General Atomics) reactor as a 

neutron source for performing neutron radiography irradiations on small test components. By the end 

of 1979, NRAD was only equipped with a single beam tube; however, in 1982 the facility was 

expanded outside the Service Floor to incorporate a second beam tube. This eliminated the need for 

all test components to pass through the HFEF-N’s Main Cell, which was significantly contaminated 

with alpha radioactivity (HFEF FSAR, 1977; NRAD Upgrade, 2014; North Radiography Station, 

1982).
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The NRAD reactor is a portion of the TRIGA-conversion-type reactor originally located at the Puerto 

Rico Nuclear Center (PRNC) (HFEF FSAR, 1977; NRAD Upgrade, 2014). In 1971, the PRNC 

Rector was converted to a TRIGA-FLIP (Fuel Life Improvement Program) system that used 70% 

enriched uranium (HFEF FSAR, 1977; NRAD Upgrade, 2014). In September 1976, the PRNC 

Reactor was shut down (HFEF FSAR, 1977; NRAD Upgrade, 2014). Then in 1977, a portion of the 

PRNC reactor fuel elements and other components (e.g., control rods, control-rod drives, and 

instrumentation/control console) were moved by the DOE to the ANL-W site and built into the NRAD 

Reactor (HFEF FSAR, 1977; NRAD Upgrade, 2014). The NRAD Reactor first went critical in 

October 1977 (NRAD Upgrade, 2014); it became operational in 1978 (HFEF History, 1979; NRAD 

Upgrade, 2014). 

NRAD Facility is primarily used for neutron radiography analysis of both irradiated and unirradiated 

nuclear fuels and materials. Typical applications for examining the internal features of nuclear fuel 

elements and assemblies include: fuel pellet separations, fuel central-void formation, pellet cracking, 

evidence of fuel melting, and material integrity under normal and extreme conditions. Examination of 

the behavior of large test loops and assemblies can also be performed (NRAD Upgrade, 2014).

Truck Lock

The Truck Lock is the primary location where casks and equipment are transported to and from the 

HFEF-N (HFEF Safety Report, 1975). 

Cask Tunnel 

The 70 ft. long x 12 ft. wide x 13 ft. high Cask Tunnel is located on the Service Floor level and 

extends beneath the Truck Lock, Decon Cell, and a portion of the Main Cell (HFEF Design, 1976). 

The Cask Tunnel is used for transferring equipment and irradiated components to and from the Truck 

Lock to the Decon Cell and/or Main Cell. 

Partition walls and controlled air-flow patterns between the various Service Floor areas are used to 

help prevent the spread of radioactive contaminants in the event of an accidental release (HFEF 

Design, 1976). 

Transfer Tunnel 

The heavily-shielded Transfer Tunnel is located on the Service Floor level and extends beneath the 

Decon Cell and a portion of the Main Cell (HFEF Safety Report, 1975; HFEF Design, 1976). The 

shorter Transfer Tunnel runs parallel to the Cask Tunnel; the two tunnels are separated by a thick 

concrete shield wall. The Transfer Tunnel is used for the movement of large equipment and irradiated 

components between the Decon Cell and Main Cell (HFEF Safety Report, 1975). 

On the west end of the Transfer Tunnel, there is an entrance that permits infrequent personnel entry. 

This entry way is normally closed by inner steel doors and a rectangular shielding plug. This door is 

supported on air-bearing support pads and is manually opened and closed (HFEF Safety Report, 

1975). When closed, the inner steel doors are sealed with duct tape to prevent outleakage of 

radioactive particulates in case the tunnel air pressure inadvertently becomes positive relative to other 

parts of the Service Floor (HFEF Safety Report, 1975; HFEF Design, 1976). 
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HFEF-N Operating History

HFEF-N construction was completed in 1972 (Site Plan, 1981). Documents indicate that HFEF-N 

was not activated until March 1975 (HFEF History, 1979; Bacca, 1985; Site Plan, 1981). However, 

survey records indicate that radiological work was initiated at HFEF-N well before March 1975. By 

August 1972, routine surveys of HFEF-N had been initiated; however, there is no indication of 

radiological work being performed until October 1972 (HFEF Survey Data, 1972-1973, PDF p. 14). 

A survey record indicates that grinding was performed on depleted uranium in October 1972 (HFEF 

Survey Data, 1972-1973, PDF p. 49); this was likely done outside the hot cells. A Special Survey 

Report indicates that “Pu oxide element E-6” was transferred into the Decon Cell on December 21, 

1972 (HFEF Survey Data, 1972-1973, PDF p. 32). 

In August 1976, a highly-sensitive fixed air-sampling system (FASS) was put into operation at 

HFEF-N. The system consists of 66 sampling stations dispersed throughout the facility to provide 

data on any radioactive particulates in the atmosphere. The sensitivity of the FASS enabled early 

detection of airborne alpha radioactivity in concentrations as low as 1 x 10-15 μCi/cm3. The FASS 

data would alert operation personnel to airborne contamination problem and reflect the long-term 

trend in the effectiveness of contamination controls (HFEF History, 1979; Courtney, 1977).

The NRAD Reactor first went critical in October 1977 (NRAD Upgrade, 2014); the NRAD Facility 

became operational in early 1978 (HFEF History, 1979; NRAD Upgrade, 2014).

Between 1977 and 1979, the Hot Repair Area was extensively modified and expanded to facilitate the 

efficient repair of manipulators and other in-cell equipment, and to enhance the controls for 

preventing radioactive contamination of the High-Bay Area. The modifications to the Hot Repair 

Area provided (HFEF History, 1979): 

• readily-decontaminated surfaces

• a two-station shielded glove-wall equipment repair area

• a glove box system for decontamination and repair of non-sealed master-slave manipulators 

• a glove box system for repair of sealed manipulator slaves

• a buffer zone between the Hot Repair Area and the remainder of the High-Bay Area 

5.1.2.7 Zero Power Plutonium Reactor (ZPPR) 

ATTRIBUTION: Section 5.1.2.7 was completed by Jason Davis, Oak Ridge Associated 

Universities. All conclusions drawn from the data regarding ZPPR were peer-reviewed by the 

individuals listed on the cover page. The rationale for all conclusions in subsequent sections of this 

document devoted to ZPPR are explained in the associated text. 

The Zero Power Plutonium Reactor (ZPPR) facility was a research facility designed to test the 

neutronics of' large, fast-neutron power reactors of' many different sizes, compositions, and shapes. 

ZPPR helped define many of today’s nuclear reactor core designs through the use of standard, 2.54 

cm-wide nuclear (e.g., up to DOE Category I) and non-nuclear materials. Many of these materials are 

still stored at the facility.
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Source: ZPPR, undated

Figure 5-42: ZPPR Reactor and Support Areas

The reactor assembly itself' was housed in a cylindrical room with a reinforced concrete wall 

surrounded by earth fill (see Figure 5-42). The roof structure consisted of steel cables, wire mesh, and 

earth and gravel fill in the form of an inverted cone. The vault and work room, together with some 

portions of the control room, were located in and under the earth fill surrounding the assembly room 

(see Figure 5-43). The various work areas, control room, vault, work room, assembly room cell, were

connected by tunnel corridors equipped with blast absorbers, blast- and gas-tight doors, and fire doors 

(ZPPR, 1963).
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Source: ZPPR, 1984

Figure 5-43: ZPPR Vault Workroom Area

The critical assemblies (reactors) were constructed within a large stainless-steel honeycomb structure 

called the ZPPR matrix consisting of two halves, each 14 ft. x 14 ft. x 5 ft. When the reactor was shut 

down, the halves were separated by a large gap. This feature, known as a split-table design, allowed 

easy access to the very center of the core, while at the same time providing a large safety margin for 

personnel operations within the cell. The matrix halves were driven together by three electric motors, 

and then the reactor was operated normally by remotely-driven control rods (ZPPR, 1985). 

The ZPPR fuel elements were called "drawers.” Drawers were 2 in. x 2 in. x 3 ft. long stainless-steel 

containers loaded with rectangular pieces of reactor materials. These pieces were 2-in. high, from 

1 in. to 12 in. long, and varied from 1/16 in. to 2 in. wide. The sodium coolant and the 

plutonium/uranium fuels were clad in stainless-steel containers (ZPPR, 1985).

Authorization to operate ZPPR was granted on March 27, 1969. Initial criticality was achieved on 

April 18, 1969, with a reactor loading in the FTR-1 pattern having a total fissile mass (Pu-239, Pu-241 

and U-235) of 362.840 kg. In January 1970, loading of ZPPR Assembly 2 began. This assembly was 

a demonstration-plant benchmark reactor in a size projected for 350-500 MWe. Initial criticality was 

reached in March 1970 with a volume of 2398 liters containing 1,066 kg Pu-239 and Pu-241 (ZPPR, 

1970).
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Any change in the composition and/or configuration of the ZPPR critical assembly was documented 

in a written Loading Change Plan. The Loading Change Plan contained a detailed description of the 

loading of each drawer or calandria involved in the change, as well as a detailed plan for the changes 

made in the reactor matrix loading. Forms were also expected to include an estimate of the reactivity 

change that would have resulted from the loading change. All personnel involved in work activities in 

the mound area wore, at a minimum, lab coats, gloves, and shoe covers. All personnel within the 

reactor cell during fuel unloading operations, or any person working in a reactor plenum was required 

to wear a full-face respirator (ZPPR, 1984).

Fuel plates or rods were normally loaded into pre-loaded drawers or calandria. All normal loading 

changes were done in the workroom, and the loading of plutonium fuel plates into drawers and 

plutonium fuel rods into calandria cans were done in the loading hoods in the workroom by gloved 

personnel (ZPPR, 1984). 

Fissile materials were kept in the vault when they were not being loaded or in the reactor cell. Each 

plutonium plate removed from the canisters was smeared for alpha contamination (at least one large 

surface and both closure welds per plate), and the results of these smears had to be available before 

loading the affected plates into drawers. The plates were allowed to be batch-smeared, but if a single 

batch-smear indicated more than an average of 25 dpm per plate, the plates were to be smeared again 

individually (all surfaces). Plates exceeding 25 dpm on the individual smears were considered 

contaminated, placed in a fuel pot, and returned to the Special Materials (SPM) representative (ZPPR, 

1984). 

After smearing, the fuel plates were passed through a go/no-go gauge and, if a plate was found to be 

oversized or otherwise deformed, it was placed in a fuel pot and returned to the SPM representative 

(ZPPR, 1984).

The loaded drawers and calandria cans could be transferred to the reactor in two ways: (1) the drawers 

were removed from the hood and carried to the transfer tube on a drawer cart, then delivered to the 

reactor cell via the transfer tube, and then transferred to the reactor via another drawer cart; or (2) the 

drawers were removed from the hood and placed on a drawer cart, then the cart would be pushed into 

the reactor cell via the access corridor, and then the drawers placed in the reactor. For safety, two 

full-time ANL-W employees were required in the reactor cell whenever the cell was occupied (ZPPR, 

1984).

Fuel unloading was similarly regimented. At least two people in full-face respirators were required to 

be in the cell during any unloading operations. An alpha probe was used in the cell for initial PuO2 

surveillance, and neutron shields would be used in such a way that the person between the halves 

would be exposed to as little of the core surface area as possible. One of the two workers would go 

between the halves, pull a drawer carefully and expeditiously, and place it on a drawer cart or pass it 

to another worker to place on a drawer cart. The Health and Safety (HS) representative standing

nearby would first survey each drawer for Pu alpha contamination as it was placed on the drawer cart 

using an alpha probe meter. When the cart was filled, they would smear the top of the fuel column 

and count the smear before the cart was allowed to leave the cell. The personnel in the cell were 

prohibited from removing their respirators until they were told that the smear indicated that there was 

no loose plutonium (<25 dpm/plate). The fuel was subsequently smeared in the workroom. If the 
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smear taken in the workroom indicated that there was loose plutonium in a drawer, the cell personnel 

were to be informed and were to evacuate the cell (ZPPR, 1984).

To transport drawers to the workroom via the conveyor, the tube door in the reactor cell was opened, 

the drawer inserted, and the conveyor advanced. No more than 10 fuel drawers or 13.0 kg of Pu were 

allowed on the conveyor. At the workroom end, after the cell tube door was closed, the workroom 

door was opened and the drawers removed. The cell and workroom tube doors were interlocked to 

prevent simultaneous opening. Similarly, if the drawer carts were used, no more than 10 fuel drawers 

or 13 .0 kg of Pu were allowed on one cart. The carts were wheeled from the cell through the access 

corridor and placed near the loading hoods (ZPPR, 1984).

Plutonium plates were to be removed from the drawer and batch-smeared (at least one large surface 

and both closure welds per plate). If the batch-smear count indicated less than an average of 25 dpm 

per plate, the diluent materials and the drawer were assumed uncontaminated and removed from the 

hood and placed into appropriate storage bins. If the batch-smear indicated more than an average of 

25 dpm per plate, the plutonium plates would be individually smeared and the diluent columns 

adjacent to the fuel were batch-smeared. All plutonium plates with smear counts greater than 25 dpm 

were put into fuel canisters and returned to SPM for further disposition. If the diluent-material smears 

were less than 200 dpm/U3O8 plate, or less than 100 dpm/other plates, all of the diluent materials from 

the drawer were permitted to be removed from the hood and placed into the appropriate (usable 

material) storage bins; if not, the smeared materials were put into a plastic bag. The bag was sealed 

and put into a bin designated for contaminated diluent materials. Drawers that contained 

contaminated materials were bagged and taken out of the hood and put into appropriate storage 

containers (ZPPR, 1984).

Contamination control was a primary focus during ZPPR loading and loading operations. Smearable 

alpha activity on the working surfaces of the hoods was required to be less than 10 dpm/100 cm2

before loading or unloading operations could begin. Similarly, smearable alpha-particle activity on 

working surfaces of tables was limited to less than 25 dpm/100 cm2

. A Health and Safety (HS) 

representative was required to smear each hood and table to assure conformance with these levels. If 

the contamination level of a hood or table exceeded the control level, it was to be cleaned until that 

level was achieved. The hoods and tables were to be re-checked for contamination periodically (at 

least twice per day) during loading and unloading operations in which they were involved (ZPPR, 

1984). 

Floors and exposed surfaces in the reactor cell and/or workroom were cleaned prior to the beginning 

of loading or unloading operations whenever the contamination level was above 25 dpm/100 cm2

. At 

least five smears were used to establish contamination levels in the cell at the following locations: (1) 

on the loading platform; (2) in the plenum behind Half 1; (3) in the plenum behind Half 2; (4) on the 

floor along the path of travel from the matrix to the transfer port; and (5) on the floor along the path of 

travel from the matrix to Door 70. At least three smears were to be obtained in the workroom at the 

following locations: (1) on the floor along the path of travel from the conveyor tube to the loading 

hood; (2) on the floor along the path of travel from Door #67 to the loading hood; and (3) on the floor 

along the path of travel from the loading hood to the utility hood and vault (ZPPR, 1984).
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Irradiation of hundreds of plutonium and uranium foils was a standard ZPPR experiment for 

determining detailed reaction rate and power distributions. The irradiated foils were analyzed in the 

ZPPR gamma and alpha spectroscopy laboratory situated in the basement of the office complex. The 

laboratory was also equipped with an alpha-counting system, used mostly for determination of 

absolute masses on foil deposits (ZPPR, 1985).

5.1.2.8 Laboratory & Operations Building (L&O Building) 

ATTRIBUTION: Section 5.1.2.8 was completed by Jason Davis, Oak Ridge Associated 

Universities. All conclusions drawn from the data regarding the L&O Building were peer-reviewed 

by the individuals listed on the cover page. The rationale for all conclusions in subsequent sections of 

this document devoted to the L&O Building are explained in the associated text. 

On June 23, 1959, Argonne National Laboratory moved its ANL-W offices from the EBR-I Complex 

to the Laboratory & Services Building (later renamed L&O Building) in the EBR-II Complex. This 

was the first occupancy at the EBR-II Area (EBR-II, undated). The L&O Building, including the 

Analytical Laboratory (AL), was built in the late 1950s and was operational in the early 1960s. 

Important functions of the analytical laboratory staff included assisting reactor personnel in the 

detection of fuel-element failures in the reactor, and locating the failed element or elements in a 

subassembly in the core (EBR-II Fuel Cycle Story, 1987).

Analytical and control support activities were conducted at the L&O Building. Six analytical caves 

(hot cells) were provided to handle small radioactive samples, including irradiated fuel samples and 

radioactive samples of the fuel elements and process materials. These samples included fuel alloy, 

cladding, oxidized skull, and scrap. Analytical facilities were also provided to support operation and 

control of argon and sodium systems, and a variety of waste processes (EBR-II, undated). The work 

initially performed in the hot cells supported the recycle of EBR-II fuel during the late 1960s. 

There were three wings on the main floor of the L&O Building: a junior cave wing, a conventional 

laboratory wing, and an office wing (EBR-II Fuel Cycle Story, 1987). The AL occupies the main 

floor of the north wings (A & B Wings) and includes the Nondestructive Analysis (NDA) Laboratory 

located at the east end of B Wing. The AL also included two counting rooms and a storage room in 

the B Wing basement, in addition to the ventilation exhaust filter areas in the A and B Wings of the 

basement. The AL complex consists of the following: 

A Wing: 

• Six shielded hot cells and attached gloveboxes

• A decontamination and manipulator repair room 

• Two storage vaults 

• One general chemistry/instrumental laboratory

• Gloveboxes and analytical instruments 
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B Wing: 

• Eight general chemistry laboratories

• Analytical glovebox laboratories 

• Mass spectrometry laboratories

• Three counting rooms (two in the basement) 

• One chemical storage room

• One experimental fuel casting laboratory

• An NDA laboratory

• Basement utility area, including B-50/51 

Source: EBR-II Fuel Cycle Story, 1987

Figure 5-44: The Analytical Laboratory Wing of the L&O Building

The cave wing was divided lengthwise by the block of caves into an Operating Area and a Service 

Area (see Figure 5-44). Another full-length partition separated the cave area from an area containing 

the upper part of a waste evaporator, storage vaults for special nuclear material, and a 

decontamination area. Transfer locks for introducing materials to, and removing materials from, the 

caves are located in the back walls of Caves 1 and 6. A hot-drain system was provided with openings 

into the system at one point on each work tray with a header leading to a hot-waste tank adjacent to 

the solidification system in the basement area. However, shortly after work with highly-active 

samples began, it was determined that the wastes were likely to be too hot to handle in this manner. 

Since then, highly-active liquid wastes were sorbed on vermiculite in plastic bottles for pneumatic 

transfer to a shielded cask for disposal (EBR-II Fuel Cycle Story, 1987).
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An important accessory to the junior caves was a decontamination room located adjacent to the 

Service Area. This room was used to reduce the level of activity from malfunctioning cave 

accessories in order to repair or modify them, as well as to keep the surface activity levels of transfer 

casks low.

The cells ("caves") of the analytical junior cave facility were designed to safely handle analytical 

samples with a radioactivity level of 20 Ci of 1-MeV gamma radiation (see Figure 5-45). The basic 

purpose of the hot cells was to provide a shielded laboratory environment for the analysis of nuclear 

fuel and cladding samples that were irradiated in EBR-II. In addition, the laboratory was intended to 

handle most analytical problems occurring during EBR-II operation. Each hot cell contained one 

26-inch-thick leaded glass shielding window. The window contained three seven-inch-thick panes of 

leaded glass cover plates. The leaded glass panes were separated from one another by a void 

containing mineral oil. Each window weighed approximately 5500 lbs. The walls separating the hot 

cells from one another had a 1 ft. by 2 ft. hole through which passed a transfer system. The purpose of 

the transfer system was to transfer tools, samples, chemicals, and other support equipment from one 

cell to another (ANL, undated). 

Source: Michelbacher, 1995

Figure 5-45: The Analytical Laboratory Junior Caves

Fuel was transferred from one cell to another with the aid of the transfer system located in front of the 

cells. The transfer system consisted of a cart that rode on a stainless-steel conveyor tray supported by 

a steel stanchion. The cart held the transfer materials and was pushed from cell to cell by the 

master/slave manipulators (ANL-W S&H Plan, 1992).

Each hot cell had a unique function for the analysis of nuclear fuel samples. Cell #1 was the 

waste-compacting and disposal cell. The cell contained a storage rack for fuel-pin storage, and the 

floor sump received contaminated liquid from floor drains in other cells. Cell #2 was the fuel 

dissolution cell. Dissolved fuel was placed in containers and transferred to Cells #3, #4, #5, or #6 for 

a specific type of analysis. Cell #2 was also used for waste reduction of liquid fuel samples. Within 

Cell #3, analytical techniques such as retained fission gas analysis, radiochemistry analysis, sample 

burn-up analysis, and fuel preparation were performed. A 1000 lb. capacity monorail crane was 
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installed in Cell #4. Physical characteristics of fuel were analyzed in this cell. Atomic absorption of 

fuel samples was performed in Cell #5.

From the start of fuel recycling operations in 1964 through most of 1968, all unclad fuel samples from 

the Argon Cell were transferred (in groups of up to six samples) in capped 0.75-in. by 2.5-in. 

aluminum cans held in holes bored into a hexagonal aluminum block. The blocks were moved 

through a transfer lock into the Air Cell transfer port for insertion into the cavity of a top-loading cask 

for shipment to the Junior Caves of the Hot Analytical Laboratory. These operations were frequently 

accompanied by the dispersal of contaminated particles on the upper surfaces of the cask, within the 

cavity, on the outer wall of the cell, and on the floor beneath the port. Particle dispersal was reduced 

somewhat by taping plastic around the lower part of the cask. However, it was necessary to remove 

the plastic in order to lower the sample block out of the cask and place it into the breach-loading 

transfer device of the Junior Cave. This resulted in the spread of contamination at that point. The 

contamination problem was resolved in the latter part of 1968 by the construction of a pneumatic tube 

transfer system from the Air Cell and its subcell to the Junior Cave system. Individual sample cans 

were transferred by this system rather than the blocks of six samples, and worker exposure during 

these processes was virtually eliminated (EBR-II Fuel Cycle Story, 1987).

Conventional wet chemical laboratories with hoods, in conjunction with the junior caves, were 

required for several functions: the preparation of reagent solutions and analyses of nonradioactive 

materials; the development of methods for the analysis of both radioactive and nonradioactive 

materials; the conduct of routine and special analyses when massive shielding was not required; and 

for miscellaneous investigations (since there was little other work space available). Two such 

laboratories (20 ft. x 25 ft.) were initially provided. They were equipped with standard laboratory 

benches on the side walls and a double-hood island in the middle. Two other substantially-identical 

rooms were also available, but were initially used as offices for FCF and laboratory staff. As other 

office space was obtained and laboratory needs increased, these two rooms were converted into a third 

wet chemical laboratory and a laboratory provided with an inert-atmosphere (argon) and an alpha (airatmosphere) glove box. 

The AL’s mass spectrometer was installed in September 1960. The spectrometer provided isotopic 

analyses of uranium, plutonium and other elements, as required the FCF (Reactor Operation, 1962). 

Burn-up measurements were made on irradiated fuel using isotopic dilution mass spectrometry 

techniques to measure the fission products La-139 and Nd-148. Because isotopes of xenon in various 

proportions were added as identification tags to experimental capsules irradiated in EBR-II, a gas 

mass spectrometer was also installed in this laboratory for reactor gas analyses to identify leaking 

experimental subassemblies (Stevenson, 1987). 

A small stockroom for laboratory supplies and equipment had been provided adjacent to the mass 

spectrometer laboratory. Around 1962, it was determined that laboratory space was needed for 

preparing specimens of structural materials and unirradiated fuel for metallographic examination to 

resolve certain fuel fabrication and performance problems. Consequently, the stockroom became a 

metallurgical laboratory. In 1971, when better facilities for that purpose became available in the Hot 

Fuel Examination Facility/North, the met lab was equipped for preparing gas mixtures for the 

xenon-tag program. 
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In 1971, after a larger machine shop was provided in another building, the L&O Building machine 

shop was converted to office space. A 12 ft. long section of this area was partitioned off and used as 

an instrument laboratory associated with junior cave work (EBR-II Fuel Cycle Story, 1987). The 

L&O Building also contained a number of tenant activities, including major office areas, a cafeteria, 

and a library, none of which contained radioactive or extremely hazardous substances.

5.1.2.9 Argonne Fast Source Reactor (AFSR) (2nd location): 1970-1997 

ATTRIBUTION: Section 5.1.2.9 was completed by Jason Davis, Oak Ridge Associated 

Universities. All conclusions drawn from the data regarding AFSR (second location) were peerreviewed by the individuals listed on the cover page. The rationale for all conclusions in subsequent 

sections of this document devoted to AFSR (second location) are explained in the associated text. 

In the fall of 1970, the AFSR reactor was moved to a new location adjacent to the ZPPR facility at the 

ANL East Area site of the NRTS. In the course of (and as a result of) this relocation, some 

modifications were undertaken for both the reactor and reactor building. Within a 55-meter radius of 

the new reactor location were located all of the ZPPR facilities and up to 75 personnel. The Lab and 

Office Building and EBR-II facilities fell within a 300-meter radius. Approximately 550 personnel 

were employed in this area.

Prior to the move, AFSR was housed in a Butler-type building with a self-contained ventilation 

system. All building exits were to the outdoors, and there were no ventilation ductwork connections 

with adjacent facilities. Subsequent to the relocation, AFSR was housed in a portion of the ZPPR 

support wing. Conditioned air was supplied to the AFSR room at 4800 cfm from the ZPPR 

service-wing ventilation system. This air exhausted in part (4600 cfm) to the outside via two exhaust 

fans, one each on the east and west walls. The reactor-pit-ventilating air (280 cfm) was drawn from 

the AFSR room through the reactor pit and through the blower pit. Since this was a negative-pressure 

system, airborne radioactive material in the reactor pit or blower pit would be expected to have been 

retained in the exhaust air system and removed by the HEPA filters.

The AFSR reactor-cooling system drew air from the room. The flow path was around the outside of 

the depleted-uranium blanket and around the outside of the core. The reactor-cooling blower drew 

this air through a filter and exhausted it through a 30-ft-high stack on the east wall of the building. 

This was also a negative-pressure system before the blower and airborne radioactive material 

originating in the blanket core region would have been removed in the filter system. 

The essential differences between the reactor shields at the two locations are that the shield at the new 

location was straight-sided from top to bottom, in contrast to the pyramidal shape of AFSR West. In 

addition, the AFSR’s cylindrical stepped beam-hole-plug and hole was replaced during relocation 

with a four-foot-square hole and rollaway concrete plug. The interface between the inner end of the

rollaway-plug hole and the depleted-uranium blanket was covered with a flat 1/4- in. thick aluminum 

plate on the reactor after relocation (Matlock, 1976).
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5.1.2.10 Miscellaneous Facilities

ATTRIBUTION: Section 5.1.2.10 was completed by Mitch Findley, MJW Corporation. All 

conclusions drawn from the data regarding Miscellaneous Facilities were peer-reviewed by the 

individuals listed on the cover page. The rationale for all conclusions in subsequent sections of this 

document devoted to Miscellaneous Facilities are explained in the associated text.

In addition to the primary radiological facilities at the EBR-II complex, there were several 

miscellaneous support facilities of importance. The Radioactive Scrap and Waste Facility served 

primarily as a repository for retrievable scrap generated from ANL-W operations, primarily EBR-II. 

The Radioactive Liquid Waste Treatment Facility did not begin radiological operations until after the 

end date for the evaluation period for this report. The Industrial Waste Pond (now used for industrial 

water and storm drainage) has never contained radioactive material.

Radioactive Scrap and Waste Facility

The ANL-W Radioactive Scrap and Waste Facility (RSWF) is a subsurface dry storage facility 

located approximately one-half mile north of the EBR-II complex, as shown in Figure 5-46). It 

consists of a four-acre fenced area that employs a grid of vertical storage holes lined with steel liners 

for storage of solid wastes and scrap from ANL-W activities. There are no permanent buildings 

located at the RSWF. When it was constructed in 1965, the elevation of the facility was set above the 

surrounding land by adding several feet of earth. This was done to preclude flooding by surface 

runoff that had previously caused problems at the INL Burial Ground (Witbeck, 1976).

RSWF’s initial purpose was to store highly-radioactive solid waste generated primarily from the 

EBR-II fuel-refining operations performed at the Fuel Cycle Facility. The first waste disposal in the 

facility was reported in a 1965 Reactor Development Program Report, which stated: “The use of the 

EBR-II burial ground for waste has been initiated with the disposal of two cans of waste. One can 

contained an estimated 1500 curies and showed a radiation level of 80 R at 1 ft.; the other contained 

an estimated 6500 curies and read 250 R at 1 ft.” (Progress Report, Jun1965). Later, irradiated fuel 

element and subassemblies, and other irradiated materials associated with the Liquid Metal Fast 

Breeder Reactor program, were also stored at the RSWF. The primary advantage to this type of 

storage is that, if in the future it becomes desirable to do so, it would be possible to retrieve any 

individual can of scrap. (Radioactive scrap is material that contains enough source or special nuclear 

materials to be economically recoverable at some time.) The facility has expanded over the years; in 

2001, there were 1350 liners with 939 filled to capacity. (Smith, 2001).
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Source: Site Plan, 1981

Figure 5-46: ANL-W Radioactive Scrap Waste Facility (foreground)

The storage liners at the RSWF are arranged in rows with about 40 liners per row (see Figure 5-47). 

The upper end of the liners are nominally four inches above grade. The liners are welded closed at the 

bottom end and have a top closure plate that is welded on after canned material has been deposited 

into the storage tube. The storage liners are just over 12 feet in length and are constructed of 16-inch 

OD (outer diameter) carbon steel. In addition to the 16-inch OD liners, there are also 21 liners 

constructed of 26-inch OD carbon steel. These oversized liners are used for the storage of wastes

from the Sodium Loop Safety Facility (Safety Report, 1983). The containment cans used to store 

material in the steel liners are modified carbon steel paint cans that are five feet long and 11-3/4 inch 

OD. Within each of these cans, material is contained in smaller cans. There can be up to five smaller 

cans loaded and sealed within an inert gas (argon) hot cell. The lid of each five-foot can is machinecrimped (Witbeck, 1976).
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Source: Smith, 2001

Figure 5-47: RSWF Storage Liner Array and Schematic

Transport of scrap and waste to the RSWF has always been a fairly simple procedure. The 

containment cans are loaded into a shielded cask at the Fuel Cycle Facility and transported to the 

RSWF. The bottom of the transport cask is covered during movements to and from the RSWF so that 

any contamination from inside the cask or outside the scrap or waste container cannot spread. 

Deposition of steel cans into steel liners is carried out via the shielded transport cask with a remote 

door-opening cable device that allows for continuous shielding while the containment can is lowered 

into the liner. For scrap cans, a cable is left in place for future retrieval operations. Until 1976, waste 

(but not scrap) storage liners were shielded with six feet of gravel instead of concrete plugs, and no 

retrieval cable was installed (Safety Report, 1983). Since 1977, the only materials accepted for 

storage at RSWF have been solid radioactive waste that contains metallic sodium or solid radioactive 

scrap, which may or may not contain metallic sodium. Before 1977, non-sodium-containing 

radioactive waste was stored at the RSWF. Thirty of the pre-1977 non-sodium waste-containing 

liners were retrieved from the RSWF and shipped to the INL Burial Ground (RWMC) (Safety Report, 

1983). Since 1977, all non-sodium-containing radioactive waste has been sent to the RWMC.
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The Health Physics section had the responsibility of approving entry to the RSWF, designating the 

hole location for each burial, providing constant monitoring of functions during any burial operation, 

maintaining official logs and records of the burial ground operations as necessary, and periodically 

testing holes in the burial ground for water seepage (Hogg, 1971). A 1983 report indicated that over 

700 operations have been conducted at RSWF with an average personnel exposure < 0. 01 

man-Rem/operation (Safety Report, 1983).

RSWF retrieval operations may be of several different types depending on the type of waste or scrap, 

the method of packaging at the time of storage, and the method used during storage. Retrieval 

methods may consist of: (1) removing the whole RSWF liner intact from the soil, thus using the liner 

as the waste package; (2) removing the waste or scrap can from the liner and leaving the liner in the 

soil for future use; or (3) removing the scrap EBR-II subassemblies from the liner and leaving the 

liner in the soil (Safety Report, 1983). 

Radioactive Liquid Waste Treatment Facility

The Radioactive Liquid Waste Treatment Facility (RLWTF) processes low-level radioactive liquid for 

ANL-W. The RLWTF supports EBR II, HFEF, TREAT, ZPPR, and ANL-W support facilities. 

The RLWTF is capable of evaporating approximately 60,000 gallons of radioactive liquid annually, 

with the resulting solidified residue packaged and stored in an environmentally-acceptable form for 

interim storage or shallow land burial. The RLWTF did not become operational until June 1983 and 

is therefore outside the time period under evaluation. (Black, 1984) 

Industrial Waste Pond 

The Industrial Waste Pond was subject to chemical contamination during its history but never 

contained radioactive material. Consequently, it will not be covered in this evaluation. (Personal 

Communication, 2015e) 

5.2 Internal Radiological Exposure Sources from ANL-W Site Operations

Radionuclides with the widest application throughout ANL-W facilities during the period under 

evaluation were:

• Mixed fission and activation products 

• Radioiodines (I-129, I-131, I-132, I-133, I-134, and I-135) 

• Radioactive noble gases (Ar-41, Kr-85m, Kr-87, Kr-88, Xe-129m, Xe-135, Xe-135m, and 

Xe-138) 

• Plutonium (Pu-238, Pu-239, Pu-240, Pu-241, Pu-242) 

• Uranium (U-232, U-233, U-234, U-235, U-236, U-238) 

• Thorium (Th-228, Th-230, Th-232) 
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• Other Actinides (Ac-227, Pa-231, Am-241, Am-243, Cm-242, Cm-243, Cm-244, Cm-245, 

Cm-246, Np-235, Np-237) 

• Other radionuclides (e.g., H-3, Be-10, C-14, Na-24, Al-28, Fe-55, Co-60, Zn-65, Sr-90, Zr-95, 

Nb-93, Nb-95, Ag-110, Sb-122, Sb-124, Cs-137, Cs-139, Ce-141, Ce-144, Pr-140, Tm-170, Hf175, Hf-181, Ta-179, Ta-182, W-185, W-187, Tl-204, Po-210, Ra-226, and others produced by 

reactor neutron irradiation) 

The following subsections provide an overview of the internal exposure sources for the ANL-W class 

under evaluation. 

5.2.1 Internal Radiological Exposure Sources from EBR-I Complex Operations

ATTRIBUTION: Section 5.2.1 was completed by Mike Mahathy, Oak Ridge Associated Universities, 

and Mitch Findley, MJW Corporation. All conclusions drawn from the data regarding the EBR-I 

Complex were peer-reviewed by the individuals listed on the cover page. The rationale for all 

conclusions in subsequent sections of this document devoted to the EBR-I Complex are explained in 

the associated text.

EBR-I: Radiological operations at EBR-I commenced in 1951 and concluded in 1963. Facility 

decontamination and decommissioning was completed in 1975. EBR-I was a research reactor with no 

known radiological separations work, with the exception of the work performed in the EBR-I hot cell. 

The categories of radionuclides specific to EBR-I operations are discussed in the subsections below 

and in Section 7.2. 

ZPR-III: Radiological operations at ZPR-III commenced in 1955 and concluded in 1970. ZPR-III 

decontamination and decommissioning was completed in 1975 along with EBR-I. ZPR-III was a 

research reactor with no known radiological separations work. The categories of radionuclides 

specific to ZPR-III operations are discussed in the subsections below and in Section 7.2. 

BORAX: NIOSH can describe internal exposure sources for all five BORAX reactors due to similar 

operations. Uranium-only-based cores were used in four of the reactors (BORAX-I, -II, III, and –IV. 

A core of uranium oxide-thorium oxide was used in BORAX-V. NIOSH has determined that internal 

doses from uranium and releases of fission products were plausible during the BORAX-1 destructive 

test, from excursions of the BORAX-I, -II, -III and –IV reactors, and from testing of fuel failures at 

BORAX-V, although uranium was not detected in air concentrations at BORAX-IV or from BORAXV fuel failures. NIOSH has determined that the internal dose potential during the period under 

evaluation is associated primarily with reactor operation with some potential during loading/unloading 

of fuel cores and from waste handling. The radionuclides specific to BORAX operations are 

discussed in the subsections below and in Section 7.2. 

AFSR: NIOSH has identified documentation listing specific radionuclides generated by irradiation 

and other operations at AFSR. NIOSH has determined that the internal dose potential during the 

period under evaluation is associated with irradiations, loading/unloading experiments, and 

radioactive waste disposal. The radionuclides specific to AFSR operations are discussed in the 

subsections below and in Section 7.2. 
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5.2.1.1 Fission and Activation Products 

Mixed fission and activation products have been present at the EBR-I Complex since radiological 

operations started there in 1951. Over 200 fission and activation products were produced by the 

various reactors at the EBR-I Complex and were received at the complex via shipments of radioactive 

materials (e.g., irradiated fuels, contaminated equipment). The potential for exposure to those fission 

and activation products existed throughout the EBR-I Complex. 

5.2.1.2 Radioiodines 

At the EBR-I Complex, radioiodines (e.g., I-129, I-131, I-132, I-133, I-134, and I-135) were produced 

by EBR-I and the BORAX reactors and appear to have only been released in relatively small 

quantities during the evaluation period. 

5.2.1.3 Radioactive Noble Gases

At the EBR-I Complex, radioactive noble gases (e.g., Ar-41, Kr-85m, Kr-87, Kr-88, Xe-129m, 

Xe-135, Xe-135m, and Xe-138) were produced by EBR-I and the BORAX reactors and appear to 

have been released only in relatively small quantities, with the exception of BORAX-IV after the 

failed-fuel experiment, which primarily released Xe-138 and Kr-88 (Borax-IV Test, 1959). 

Because radioactive argon, krypton, and xenon do not represent a significant internal exposure 

concern, NIOSH finds there is no need to assess the internal doses attributable to them. Therefore, 

further discussion of the internal dose contribution from noble gases for the EBR-I Complex are not 

included in this report. (ICRP 30, 1979, PDF p. 34) 

5.2.1.4 Uranium 

EBR-I: EBR-I used three uranium cores during its operating history. In 1959, it introduced a 

plutonium core. While using the plutonium core, there were tentative plans for irradiating 

encapsulated samples of U-233. Whether the experiment was actually performed is unclear, but it 

would have been another potential source of uranium exposure (EBR-1, 1959). During its entire 

operating history, EBR-I had an outer blanket of U-238 bricks as part of its primary goal of proving 

the breeding reactor concept. The exposure potential to uranium isotopes was small because the fuel 

rods and blanket bricks were encapsulated. A breech of the cladding (e.g., the ruptured brick incident 

in 1955 and the partial core meltdown in 1955) would be examples of when uranium exposure was 

most likely. 

ZPR-III: From 1955 to late-1963, ZPR-III exclusively used uranium fuel plates for reactor physics 

experiments. Even when plutonium fuel plates were introduced at ZPR-III, depleted-uranium plates 

were still used; enriched-uranium fuel plates were stored in the Special Materials vault. Depleteduranium plates were uncoated due to their very small specific activity, but they did cause minor 

contamination. This contamination was typically found on the gloved hands of workers handling the 

bare plates. The workers’ gloves would become blackened from the oxide layer of the depleted 

uranium plates rubbing off onto the gloves. The potential for uranium exposure to U-234, U-235, and 

U-238 existed throughout ZPR-III’s operating history (Fish, 1967). 
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BORAX: The potential for uranium exposure to U-234, U-235, and U-238 existed at all five BORAX 

reactors during the entire period under evaluation. However, the exposure potential was very small 

because encapsulated fuel was used. With one exception, the fuel was irradiated and either stored as 

spent fuel or shipped to the CPP at INL without opening the capsules. The exception was the 

BORAX-I destructive test. In that test, the entire reactor structure blew up, distributing fuel element 

pieces and contaminated equipment up to 300 feet from the reactor (ANL-W History, 2006). 

Although the plume traveled up to 1-1/2 miles in a southwesterly direction, results from filter-paper 

samples, fallout plates, and continuous air monitors (CAMs) detected no significant non-natural alpha 

activity (Borax HP Documents, 1959).

AFSR: The potential for uranium exposure existed at AFSR but the potential was very small because 

AFSR received encapsulated materials, irradiated them, and shipped them away from AFSR without 

opening the capsules. Uranium would include U-233, U-234, U-235, and U-238. 

Infrequently, work tables and other equipment were contaminated during the removal of uranium foils 

but those items were measured and decontaminated. One example occurred in November 1963 when 

irradiated U-235 foils were removed. A work table used in preparing the foils prior to the irradiation, 

and unloading the foils after irradiation, was found to be contaminated to 200 dpm/100 cm2 smearable 

alpha. The table was decontaminated to less than 10 dpm/100 cm2 smearable alpha (Safety Monthly 

Report, 1963, PDF p. 14). 

5.2.1.5 Plutonium 

EBR-I: EBR-I used a plutonium core (designated Mark IV) from 1959 until the facility ceased 

operations in 1963. However, throughout its entire operating history there was an outer breeder 

blanket of encapsulated U-238 bricks. Thus, there was a small potential for plutonium exposure from

a breech in the integrity of one of the outer blanket bricks. In fact, there was a breach in the integrity 

of a blanket brick in April 1955 that precipitated in-vitro plutonium bioassay for workers with an 

elevated exposure potential. That scenario would have represented the only real plutonium exposure 

potential because there was never an integrity problem with the cladding on the plutonium core 

(fabricated and canned at Rocky Flats) (Fuel Elements, 1956). 

ZPR-III: Experience at ZPR-III with receiving and handling plutonium fuel plates began in 1960. The 

first batch of plutonium fuel plates for ZPR-III were found to be contaminated during the receipt 

survey and were sent back to ANL-E (Lemont, IL) without being placed into service. There were no 

other known problems with receiving contaminated plates after that initial event. The first full-scale 

plutonium loading of ZPR-III occurred in late 1963. Thus, plutonium exposure potential existed from 

1960 until 1969, when the plutonium fuel was shipped to ZPPR (Fish, 1967). The plutonium 

exposure potential was minimal because the fuel plates were clad in stainless steel and there was no 

known problem with cladding integrity (Personal Communication, 2015c). 
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5.2.1.6 Thorium 

BORAX: The potential for thorium exposure existed during BORAX-IV operations from 1957 

through 1959. However, only encapsulated ceramic fuels of UO2-ThO2 were used. Those elements 

(including the defective elements irradiated in 1958) were manufactured off site. Fuel elements were 

not opened at ANL-W, although some were shipped to CPP at INL for processing. Alpha monitoring 

conducted at BORAX-IV did not detect thorium contamination. 

5.2.2 Internal Radiological Exposure Sources from EBR-II Complex Operations

ATTRIBUTION: Section 5.2.2 was completed by Jason Davis, Oak Ridge Associated Universities, and 

Brian Gleckler, Dade Moeller, Inc. All conclusions drawn from the data regarding the EBR-II 

Complex were peer-reviewed by the individuals listed on the cover page. The rationale for all 

conclusions in subsequent sections of this document devoted to the EBR-II Complex are explained in 

the associated text.

Radiological operations at the EBR-II Complex have taken place since 1959. The EBR-II Complex 

was primarily an experimental nuclear reactors operations area, but also had radiological facilities for: 

reprocessing and refabricating irradiated nuclear fuel (Hot-Line fuel production), un-irradiated fuel 

production (Cold-Line fuel production), processing and examining irradiated nuclear fuels, neutron 

and photon radiography, performing laboratory analyses, machine shops, other research and support 

facilities, radioactive waste treatment and disposal, and other research and support facilities. 

The following categories of radionuclides that were specific to EBR-II Complex operations are 

discussed in the subsections below and in Section 7.2: 

• Mixed fission and activation products 

• Radioiodines (I-129, I-131, I-132, I-133, I-134, and I-135) 

• Radioactive noble gases (Ar-41, Kr-85m, Kr-87, Kr-88, Xe-129m, Xe-135, Xe-135m, and 

Xe-138) 

• Plutonium (Pu-238, Pu-239, Pu-240, Pu-241, Pu-242) 

• Uranium (U-232, U-233, U-234, U-235, U-236, U-238) 

• Thorium (Th-228, Th-230, Th-232) 

• Other Actinides (Ac-227, Pa-231, Am-241, Am-243, Cm-242, Cm-243, Cm-244, Cm-245, 

Cm-246, Np-235, Np-237) 

• Other radionuclides (e.g., H-3, C-14, Na-22, Na-24, and others produced by reactor neutron 

irradiation) 
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5.2.2.1 Fission and Activation Products 

Mixed fission and activation products have been present at the EBR-II Complex since radiological 

operations started there in 1959. Over 200 fission and activation products were produced by the 

various reactors at the EBR-II Complex and were received at the EBR-II Complex via shipments of 

radioactive materials (e.g., irradiated fuels, contaminated equipment). The potential for exposure to 

those fission and activation products existed throughout the EBR-II Complex. 

Because radioiodines and radioactive noble gases were periodically present at the EBR-II Complex in 

significant quantities without any other fission and activation products being present, these fission 

products are addressed in their own subsections below. 

5.2.2.2 Radioiodines 

At the EBR-II Complex, radioiodines (e.g., I-129, I-131, I-132, I-133, I-134, and I-135) were 

produced by the various reactors and, with the exception of the EBR-II Reactor and the FCF, appear 

to have only been released in relatively-small quantities during the evaluation period. Radioiodines 

were periodically released from failed fuel in quantities large enough to cause the EBR-II reactor 

building to be evacuated. Much larger quantities of radioiodines were released from the fuel 

reprocessing activities at the FCF. However, the FCF had gaseous effluent control systems in place to 

safely collect and store the radioiodines that were released from fuel reprocessing. Those systems 

could later release the radioiodines (which would have gone through some radioactive decay) through 

a 200-ft-stack when meteorological conditions allowed for their safe release. 

5.2.2.3 Radioactive Noble Gases

At the EBR-II Complex, radioactive noble gases (e.g., Ar-41, Kr-85m, Kr-87, Kr-88, Xe-129m, 

Xe-135, Xe-135m, and Xe-138) were produced by the various reactors and, with the exception of the 

EBR-II Reactor and the FCF, appear to have been released only in relatively small quantities. 

Radioactive noble gases were periodically released from failed fuel in quantities large enough to cause 

the EBR-II Reactor building to be evacuated. Beginning in 1972, some of the noble gas releases from 

the EBR-II reactor included isotopes of radioactive xenon that were used to tag the fuel elements so

that a breached fuel element could be readily-identified while still in the reactor (Smith, 1977). Much 

larger quantities of radioactive noble gases were released from the fuel-reprocessing activities at the 

FCF. However, the FCF had gaseous effluent control systems in place to safely collect and store the 

radioactive noble gases that were released from fuel reprocessing. Those systems could later release 

the radioactive noble gases (which would have gone through some radioactive decay) through a 200-

ft-stack when meteorological conditions allowed for their safe release. 

Because radioactive argon, krypton, and xenon do not represent a significant internal exposure 

concern, NIOSH finds there is no need to assess the internal doses attributable to them. Therefore, 

further discussion of the internal dose contribution from noble gases are not included in this report 

(ICRP 30, 1979, PDF p. 34). 
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5.2.2.4 Plutonium 

At differing times, plutonium has been present at the facilities comprising the EBR-II Complex. 

TREAT began the first irradiation of U/Pu foils in August 1959, EBR-II started up in September 

1961, FCF received its first irradiated fuel in September 1964, and L&O began analyzing irradiated 

fuels and samples in 1964. Plutonium was present on the complex for the following reasons: 

• Plutonium was produced by the various reactors at the EBR-II Complex. 

• Plutonium was used as primary and experimental reactor fuels in the EBR-II Complex reactors. 

• Plutonium was received at the EBR-II Complex via shipments of radioactive materials (e.g., 

irradiated fuels, contaminated equipment). 

• Plutonium was an impurity in the recycled uranium that was used as a reactor fuel. 

• Plutonium was present in non-dispersible forms (e.g., encapsulated sources, electro-plated 

sources).

• Plutonium may have been present as liquid laboratory standards. 

Although plutonium may have been technically present in some forms at an early date, the purpose of 

this evaluation report is to identify and evaluate both exposure potential and actual personnel 

exposure. Although plutonium was present in matrices/forms at certain locations for certain 

operations, mere presence alone does not qualify as a source of personnel exposure requiring 

accounting for the purpose of this report. 

Some sources of plutonium at the EBR-II Complex require evaluation and accounting. In such cases, 

with a few exceptions, the majority of the dispersible forms of plutonium were present with mixed 

fission products (e.g., EBR-II and FCF reactors). Plutonium without mixed fission products would 

have been contained in fresh plutonium fuels that had not aged long enough to allow for the ingrowth 

of fission or daughter products, as well as in fuels that had undergone chemical separation. The 

primary areas in which plutonium was present without mixed fission products were HFEF-N, L&O 

Building, and the TREAT and ZPPR reactors. 

At the FCF, intermittent work associated with the Mark-II type RAS-TREAT sodium-loop 

experiments could have resulted in potential intermittent exposures to plutonium without mixed 

fission products present when that work included the removal of the test sections and when that work 

was performed outside of the Air and Argon Cells. At the HFEF-N, the source of plutonium without 

the mixed fission products was also from work associated with the Mark-II type RAS-TREAT 

sodium-loop experiments. However, that work appears to have been performed exclusively in the 

HFEF-N’s hot cells, so the potential for worker exposures was minimal. At the EBR-II, TREAT, and 

ZPPR reactors, the potential for plutonium exposures was minimal because these reactors used 

encapsulated fuel sources. At the L&O Building, the primary sources of plutonium without mixed 

fission products were some of the encapsulated fuels being stored in the building as well as 

radiochemical analysis of samples containing plutonium in order to determine the breeding ratio. 

Liquid laboratory standards may have also been present in the L&O Building. 
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5.2.2.5 Uranium 

Uranium has been present at the facilities comprising the EBR-II Complex since radiological 

operations started there in 1959. Uranium was present on the complex for the following reasons: 

• Uranium was used by the reactors as fuel.

• Irradiated uranium was reprocessed and refabricated into nuclear fuel (Hot-Line fuel production). 

• Non-irradiated uranium was fabricated into nuclear fuel (Cold-Line fuel production). 

• Uranium was received at the EBR-II Complex via shipments of radioactive materials (e.g., nonirradiated fuels, irradiated fuels, contaminated equipment).

• Uranium may have been present as non-dispersible radioactive sources. 

• Uranium may have been present as liquid laboratory standards. 

The uranium at the EBR-II Complex covered the entire spectrum in regards to its U-235 content, 

ranging from depleted uranium to fully-enriched uranium. 

The majority of the dispersible forms of uranium without mixed fission products present at the EBR-II 

Complex were in the Cold-Line fuel production areas at the FCF, ITF, and FASB. At the end of 

December 1969, the Cold-Line fuel pin and fuel element production equipment in FCF Rooms 20 and 

26 was put on stand-by (Progress Report, Dec1969, PDF p. 88). However, the Cold-Line continued to 

fabricate unirradiated subassemblies from vendor-produced fuel elements (at ITF) and to perform 

impact-bonding of the unbonded vendor fuel elements (in FCF Room 23) (Progress Report, Dec1969, 

PDF p. 88). The FCF’s Cold-Line fuel pin and fuel element production was likely shut down due to a 

large backlog of completed fuel elements, and given that ANL-W was beginning to receive large 

numbers of acceptable vendor-produced fuel elements. In January 1970, there were a total of 23,766 

completed fuel elements that were ready to be put into EBR-II Reactor subassemblies (Progress 

Report, Jan1970, PDF p. 78). In the mid-1970s, the FCF’s Cold-Line fuel pin and fuel element 

production equipment was deactivated (Batte, 1986, PDF p. 9). Fuel-pin fabrication in FCF Room 20 

was not re-initiated until 1982 (Batte, 1986, PDF p. 9). Given that Cold-Line fuel production in 

FASB did not begin until February 1972, no Cold-Line fuel pin and fuel element production took 

place at ANL-W during the period of January 1970 – January 1972. 

At the HFEF-N, the source of uranium without the mixed fission products was from work associated 

with the Mark-II type RAS-TREAT sodium-loop experiments. However, that work appears to have 

been performed exclusively in the HFEF-N’s hot cells, so the potential for worker exposures was 

minimal. 

At the L&O Building, the primary dispersible sources of uranium without mixed fission products were 

the non-irradiated derbies, ingots, and fuel pins being stored in the building. Liquid laboratory 

standards may have also been present in the L&O Building. 
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5.2.2.6 Thorium 

At differing times, thorium has been present at the facilities comprising the EBR-II Complex. 

Thorium was present on the complex for the following reasons: 

• was produced in small quantities by the reactors. 

• was an impurity in the recycled uranium that was used as a reactor fuel. 

• was used as an experimental reactor fuel and/or blanket material.

• was primarily used in the form of thoria (ThO2) as part of the Hot-Line fuel production at the FCF. 

Although thorium may have been technically present in some forms at an early date, the purpose of 

this evaluation report is to identify and evaluate both exposure potential and actual personnel 

exposure. Although thorium was present in matrices/forms at certain locations for certain operations, 

mere presence alone does not qualify as a source of personnel exposure requiring accounting for the 

purpose of this report. 

Some sources of thorium at the EBR-II Complex require evaluation and accounting. Thorium that 

had undergone chemical separation and not aged long enough to allow for the ingrowth of daughter 

products would not have contained appreciable quantities of mixed fission products. The majority of 

the dispersible thorium without mixed fission products present was in FCF Room 25, which is where 

the crucibles and molds used in Hot-Line fuel production were coated with thoria. At the FCF, thoria 

was used as a coating during the period of August 1963 – November 1967. The experimental reactor 

fuels and blanket materials were encapsulated so the potential for worker exposures to that thorium 

was minimal. In the late-1970s, some contaminated thorium blanket materials were received at ZPPR; 

this is discussed in more detail below. 

In the late-1970s, the effects of substituting thorium for uranium in both axial and radial ZPPR 

blankets were studied (ZPPR, 1985). Integral experiments on ZPPR Assembly 8 were designed to test 

the effect of thorium in different blanket zones. ZPPR 8C was a uranium-plutonium oxide reactor 

with 293 kg of thorium loaded into the central blanket to simulate a thorium blanket (Thorium 

Blanket, 1978).

During the last week of September 1977, ZPPR received a 362.5 kg shipment of thorium from 

Atomics International. The thorium plates read 120 mR/hr beta-gamma at one inch when a shipping 

container was opened. Smearable activity of 880 dpm/100 cm2 alpha was detected on the plates. Five 

plates were packaged and sent to the L&O Chemistry Labs to see if a way could be found to clean the 

loose contamination from the plates.

During that same week, a shipment of thorium previously received from ANL-E was opened in a 

ZPPR hood for inspection. The material read 65 mR/hr beta-gamma at one inch and had smearable 

activity of 10,000 dpm/100 cm2 alpha. When the plates were moved for inspection, oxide could be 

seen falling off the plates. There was no personnel contamination and no activity detected outside the 

hood. The hood was effectively decontaminated, as was the shipping container, which was returned to 

storage. An air sample taken in the breathing zone during the inspection showed 3.94 x 10-12 μCi/cc 

after 52 minutes decay, or 13% of the RCG at the time for Th-232. The sample had decayed to <1% 

of RCG by the following morning. As with the Atomics International plates, five of these plates were 
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sent to the L&O Building’s chemistry labs to see if they could find an effective way to decontaminate 

and coat them so they could be used in the reactor. 

At the L&O Building, selected thorium plates were coated with Kel-F solution to determine what 

effect it would have in reducing smearable activity levels. This method reduced smearable alpha 

activity from an average of 450 dpm/plate to <10 dpm/plate. The cleaning and coating of 4283 

thorium plates was completed on November 8, 1977. There was no detectable contamination outside 

the hoods or coating machine in which all work was done. Smearable alpha activity on coated plates 

was <25 dpm/plate, which was considered acceptable for use in the reactor. The hoods and coating 

machine were decontaminated to <10 dpm/100 cm2

 alpha and <100 dpm/100 cm2 beta-gamma. There 

were no instances of personnel contamination. In all, smearable alpha activity was reduced from a 

maximum of 10,000 dpm/plate to <25 dpm/plate by this cleaning and coating technique (Weekly 

Reports, 1977). 

5.2.2.7 Other Actinides

Other actinides (i.e., excluding plutonium and uranium) have been present at the EBR-II Complex 

since radiological operations started at the EBR-II reactor in 1964. At the EBR-II Complex, other 

actinides were produced by the various EBR-II Complex reactors in small quantities via neutron 

absorption, and were received at the EBR-II Complex via shipments of radioactive materials (e.g., 

irradiated fuels, contaminated equipment). Based on the available information, the other actinides 

were never present at the EBR-II Complex without mixed fission products being present. 

5.2.2.8 Other Radionuclides

The “other radionuclides” category includes any other radionuclides that were not addressed in the 

radionuclide categories listed above. This category includes fission and activation products that had 

been chemically separated from the other mixed fission and activation products. With the exception 

of the radionuclides that are specifically addressed below, the majority of the “other radionuclides” at 

the EBR-II Complex were likely encapsulated experiments or non-dispersible radioactive sources.

Tritium (H-3) is produced by the neutron bombardment of a number of elements, and was known to 

be present in the sodium coolant of the EBR-II reactor and was likely present as a contaminant in 

some the other EBR-II Complex locations. Radioactive isotopes of sodium (Na-22 and Na-24) were 

produced by the neutron bombardment of stable sodium, and were present in the EBR-II reactor’s 

sodium coolant and in the NaK coolant used in the RAS-TREAT sodium-loop experiments. 

Carbon-14 (C-14) can be produced from a number of neutron interactions, and was likely present as a 

contaminant in carbon, nitrogen, and/or oxygen-bearing materials that were irradiated in nuclear 

reactors at the EBR-II Complex. 
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5.3 External Radiological Exposure Sources from ANL-W Site Operations

To provide consistency of radiation safety programs at the entire INL site given the large variety of 

facilities and constantly-changing contractors, the AEC-Idaho Operations Office (AEC-IDO) 

established a Health and Safety (H&S) Laboratory to provide technical support for internal and 

external dosimetry programs over the entire site, including ANL-W. The name of this organization 

was changed to Health Services Laboratory (HSL), then to the Health and Safety Division (HSD), 

then to the Idaho Center for Radiological and Environmental Sciences, and most recently, to the 

Radiological and Environmental Sciences Laboratory (RESL) (ORAUT-TKBS-0007-6). 

By policy, the INL and ANL-W sites monitored by dosimeter all personnel who were expected to 

receive any radiation dose or whose work was centered at the site (Cipperley, 1968; Cipperley, 1958). 

Dosimeters were usually stored at the operational area entrance security gates. 

Table 5-4 shows a historical summary of the beta-gamma dosimetry systems used at the INL site, and 

their corresponding Limits of Detection (LODs). 

Table 5-5 shows a historical summary of the neutron dosimetry systems used at the INL site, and their 

corresponding Limits of Detection (LODs). 
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Table 5-4: INL Site Beta-Gamma Dosimetry Systems History and LOD Summary

Period of Usea Dosimeter Exchange 

Frequency

Beta 

LODb

(mrem)

Gamma 

LODb 

(mrem)

August 1951–

March 1958

INL initial film, 552 DuPont film Weekly (n=52) 

Monthly (n=12)

30 30

August 1951–

March 1958

Reactor areas, DuPont 558 film Weekly (n=52) 30 10

March 1958–

December 1966

INL multielement DuPont 508 film Weekly (n=52)

Biweekly (n=26)

Monthly (n=12)

30 10

December 1966–

February 1974

INL multielement DuPont 508 film Weekly (n=52)

Biweekly (n=26)

Monthly (n=12)

30 10

December 1966–

February 1974

INL LiF TLD Quarterly (n=4)

Semi-ann(n=2)

Annual (n=1)

15 15

February 1974–

May 1975c

INL Atlas TLD LiF in Teflon Monthly (n=12)

Quarterly (n=4)

Semi-ann(n=2)

Annual (n=1)

30 30

December 1974-

December 1985c 

INL Harshaw two-chip TLD Monthly (n=12)

Quarterly (n=4)

Annual (n=1)

15 15

January 1986–2006 INL Panasonic four-chip TLD Monthly (n=12)

Quarterly (n=4)

15e 15d

January 1986–2006 INL Panasonic four-chip TLD Monthly (n=12)

Quarterly (n=4)

30e 10d

Source: This table is derived from Table 6-15 in ORAUT-TKBS-0007-6.

a. For many years, INL workers had a dosimeter assigned to each operating area where they worked, or they 

were issued visitor dosimetry. An exception existed from March 1970 through December 1974 when a “one 

badge, all areas” approach was used. After 1974, dosimetry reverted to the original “one badge, one area” 

system.

b. LODs are based on: Cipperley, 1958; Cipperley, 1968; Cusimano, 1963; Kalbeitzer, 1983; Gesell, 1986; 

Gesell, 1992; Perry, 1993; Ruhter, 2002.

c. ICPP began using the Harshaw TLD in December 1974, the prime contractor began in February 1975, and 

ANL-W began in May 1975.

d. The LOD was 15 mrem from January 1, 1986 to July 7 1986; 10 mrem from July 7, 1986 to September 

1989; and 15 mrem until 1993, when it returned to 10 mrem.

e. The LOD was 15 mrem from January 1, 1986 to July 7 1986 and 30 mrem after that.
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Table 5-5: INL Site Neutron Dosimetry Systems History and LOD Summary

Period of Use Dosimeter Exchange 

Frequency

LOD 

(mrem)

1951–1958 NTA Film (Kodak Type A Film) Weekly (n=52) 14a

1959–September 1976 NTA Film (Kodak Type A Film) Weekly (n=52)

Biweekly (n=26)

Monthly (n=12)

20

January 1986–2006 INL Panasonic four-chip TLD Monthly (n=12)

Quarterly (n=4)

Monthly (n=12)

15b

Source: This table derived from Table 6-16 in ORAUT-TKBS-0007-6.

a. Cipperley, 1958, PDF p. 20

b. Gesell, 1996, PDF p. 19

When addressing exposure to noble gases, external radiation dose is the overriding concern. As 

stated in ICRP 30 (when discussing internal dose from noble gases): 

Therefore, when applying the system of dose limitation described in Chapter 2 [Basic Limits for 

the Control of Internal Dose], it is clear that, for exposure by submersion in radioisotopes of the 

noble gases, external radiation will be of such overriding importance that it alone need be 

considered. (ICRP 30, 1979, PDF p. 34)

NOTE: The ICRP 30 wording above is less technical and more straightforward than later ICRP 

documents. ICRP 68 states: “For this report the behaviour of gases and vapours assumed 

in ICRP Publication 30 has been retained, but is represented using the formalism of ICRP 

Publication 66.” (ICRP 68, Section 2.3, “Dose Coefficients for the Intakes of 

Radionuclides by Workers”) 

Exposures by submersion in noble gases would be accounted for by ANL-W personnel external 

dosimetry.

The following subsections provide an overview of the external exposure sources for the ANL-W class 

under evaluation. 

5.3.1 External Radiological Exposure Sources from EBR-I Complex Operations

ATTRIBUTION: Section 5.3.1 was completed by Mike Mahathy, Oak Ridge Associated Universities, 

and Mitch Findley, MJW Corporation. All conclusions drawn from the data regarding the EBR-I 

Complex were peer-reviewed by the individuals listed on the cover page. The rationale for all 

conclusions in subsequent sections of this document devoted to the EBR-I Complex are explained in 

the associated text.

5.3.1.1 Photon 

EBR-I: Photon exposure resulted from nuclear fission as well as the decay of fission and activation 

products produced by fuel irradiation. Within the EBR-I reactor building there were hot cells, a rod 

storage “farm,” and heavily-shielded, remote-handling rooms illustrating cognizance of the 

high-energy photons present. The workplace photon energy spectra for EBR-I would have been 

dominated by high-energy photons from mixed fission and activation products. Based on review of 
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available records, the highest external exposure potential from photons was likely during the removal 

of the partially-melted core in 1956. Special shielding was installed to reduce personnel exposures 

during that special work evolution. Most of the photon dose at EBR-I would have been photons with 

energy greater than 250 keV. (ORAUT-TKBS-0007-6, PDF p. 26) 

ZPR-III: When the ZPR-III reactor was in operation, the Assembly Room was vacated to prevent 

personnel exposures. Due to the “zero power” operation of ZPR-III, there were no significant 

quantities of fission and activation products produced. Consequently, no special remote-handling 

procedures were needed at ZPR-III (Long, 1962). Photon exposure of any significance would have 

occurred while handling fuel plates. Therefore, most of the photon dose was from photons with 

energy greater than 250 keV prior to the use of plutonium fuel plates. Photon dose from photons after 

the introduction of plutonium fuel plates likely contained photon energy less than 250 keV (ORAUTTKBS-0007-6, PDF p. 26). 

BORAX: Photon exposure resulted from mixed fission and activation products and isotopes produced 

by reactors. Exposure rates varied by dates, but generally, exposure potential increased with time due 

to increased operational power and increased lengths of operation. (Low-power operations lasted only 

a few days in 1953, but there were longer-term operations at BORAX-IV and BORAX-V from 1956 

through 1958, and from 1962 through 1964.) The radiation fields at BORAX were generated 

primarily by mixed fission and activation products. Therefore, most of the photon dose has been from 

photons with energy greater than 250 keV (ORAUT-TKBS-0007-6, PDF p. 26). 

ASFR: Photon exposure was from mixed fission and activation products resulting from irradiation of 

target materials. Therefore, most of the photon dose was from photons with energy greater than 250 

keV (ORAUT-TKBS-0007-6, PDF p. 26). 

5.3.1.2 Beta

EBR-I: EBR-I operations did not have pure beta external hazards. Beta exposure would have resulted 

from mixed fission and activation products, and isotopes produced by EBR-I. The highest potential 

for beta external exposure would have been during an incident, such as the 1955 partial-core 

meltdown during which mixed fission and activation products broke through the argon cover gas and 

were released into the reactor building. Beta dose would have resulted primarily from exposure to 

electrons with greater than 15 keV (ORAUT-TKBS-0007-6, PDF p. 26). 

ZPR-III: ZPR-III operations did not have pure beta external hazards. Beta exposure would have 

resulted from the handling of uranium fuel plates. This exposure would be expected to be minimal 

due to coatings on the uranium plates. This is reflected in the discovered non-routine dosimetry 

processing results (Film Processing, 1961). Plutonium fuel plates were clad in stainless steel and 

would not be a source of beta exposure. Beta dose resulted primarily from exposure to electrons with 

energy greater than 15 keV (ORAUT-TKBS-0007-6, PDF p. 26). 

BORAX: Beta exposure resulted from mixed fission and activation products, and isotopes produced 

by reactors. Exposure rates varied by dates, but generally, exposure potential increased with time due 

to increased production and the increasing number of reactors (in 1957 and 1967). The radiation 

fields at the BORAX reactors were generated primarily by mixed fission and activation products. 

Beta dose resulted primarily from exposure to electrons with greater than 15 keV (ORAUT-TKBS0007-6, PDF p. 26). 
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ASFR: Beta exposure resulted from mixed fission and activation products, and isotopes produced by 

reactors. Beta dose resulted primarily from exposure to electrons with energy greater than 15 keV 

(ORAUT-TKBS-0007-6, PDF p. 26). 

5.3.1.3 Neutron 

Operating reactors would have been the primary source of neutron exposure during the period under 

evaluation. These reactors were experimental facilities for measuring shielding factors, reactor 

kinetics, and neutron irradiation. Thermal neutron radiation levels 10 feet from the reactor during 

operations at 100 kW were 50-1,500 n/cm2

/sec (Freund, 1960).

EBR-I: From EBR-I’s inception, shielding from fast and thermal neutrons was a consideration. 

Before operations commenced, neutron flux studies were performed at ANL in Illinois as part of the 

Experimental Breeder Reactor program. Experiments were performed to measure the anticipated flux 

at various locations in the reactor. This was done by activating gold and phosphorus foils at points in 

the reactor and along the outer surface of the uranium control blanket. These studies provided 

estimates of the resonance flux, thermal flux, thermal leakage, and equivalent fission flux when 

operating at 1 MW (Butler, 1952; Progress Report, 1953). EBR-I workers may have been exposed 

during EBR-I operation and any activities involving neutron sources. Neutron surveys were a part of 

the routine radiological monitoring at EBR-I (Survey Reports, Mar1962). A 1952 survey of the outer 

surfaces of the reactor shield concluded that the radiation levels were “generally safe with the 

shielding presently in place” (Butler, 1952). However, as a precaution, additional concrete shielding 

was added. 

ZPR-III: ZPR-III was used to mock-up reactor cores and obtain neutron physics data; thus, 

understanding the neutron flux within the reactor was of highest importance. Spectral determinations 

were made at ZPR-III as well as derived by multi-group calculations (Hess, 1960). ZPR-III Operating 

Instructions required a properly-operating neutron detector and constant monitoring of the neutron 

flux during operations, demonstrating the prominence of neutron monitoring (Operations Manual, 

1956). Neutron activity was monitored by both thermal and fast neutron detectors. With plutonium 

fuel plates, the neutron flux was a combination of spontaneous fission neutrons and neutrons resulting 

from the α,n reaction between plutonium alphas and the aluminum alloys used in the fuel trays. 

Pu-240, with two neutrons per spontaneous fission, was the principal source of neutrons from 

plutonium. Neutron exposure was considered serious enough that the neutron field between the 

halves, and near the assembly, were monitored each time a new size assembly was constructed; 

personnel exposure was limited accordingly (Long, 1962). ZPR-III workers were most likely to have 

been exposed to neutrons during fuel-plate handling. Neutron surveys were a part of the routine 

radiological monitoring at ZPR-III (Survey Reports, Mar1964). 

BORAX: BORAX reactor workers may have been exposed to neutrons from the nuclear reactors and 

from transferring neutron sources, including antimony-beryllium and polonium-beryllium. Neutron 

fields were measured Health Physics staff in the turbine building at BORAX-III, in the BORAX-IV 

reactor building, and in the BORAX-V reactor building.

ASFR: During very early operations conducted between January and June 1960, neutron shielding 

was found to be inadequate; consequently, more shielding was added (Site Plan, 1981, PDF p. 148; 

Safety Monthly Report, 1960). AFSR workers may have been exposed to neutrons from the reactor 
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and from transferring neutron sources, including polonium-beryllium. Neutron radiation was a 

concern of Health Physics staff given the use of AFSR for material irradiation through beam ports; 

therefore, neutron fields at ASFR were measured. 

5.3.2 External Radiological Exposure Sources from EBR-II Complex Operations

ATTRIBUTION: Section 5.3.2 was completed by Jason Davis, Oak Ridge Associated Universities, and 

Brian Gleckler, Dade Moeller, Inc. All conclusions drawn from the data regarding the EBR-II 

Complex were peer-reviewed by the individuals listed on the cover page. The rationale for all 

conclusions in subsequent sections of this document devoted to the EBR-II Complex are explained in 

the associated text.

5.3.2.1 Photon 

EBR-II: Based on a review of the available ANL-W records, the photon radiation fields at the EBR-II 

reactor were generated primarily by the operation of the nuclear reactor, irradiated reactor fuels, 

activated materials, and releases of radioactive noble gases. Photon energies were predominately 

greater than 30 keV (ORAUT-TKBS-0007-6). Photon radiation fields at the EBR-II reactor are 

known to have ranged from immeasurable levels to over 1 x 106 mrem/hr from irradiated fuel 

assemblies. However, the irradiated fuel assemblies were always heavily shielded inside the reactor 

pool or inside an inter-building coffin (IBC). 

FCF: Based on a review of the available ANL-W records, the photon radiation fields at the FCF were 

predominately from irradiated reactor fuels, activated materials, and un-irradiated uranium. Photon 

energies were predominately greater than 30 keV (ORAUT-TKBS-0007-6). Photon radiation fields at 

the FCF are known to have ranged from immeasurable levels to over 1 x 106 mrem/hr from irradiated 

fuel assemblies. However, the irradiated fuel assemblies emitting those radiation levels were always 

heavily shielded inside the FCF’s hot cells or inside an inter-building coffin (IBC). 

HFEF-N: Based on a review of the available ANL-W records, the photon radiation fields at HFEF-N 

were predominately from irradiated reactor fuels, activated materials, and by the operation of the 

NRAD Reactor (October 1977 and later). Photon energies were predominately greater than 30 keV 

(ORAUT-TKBS-0007-6). Photon radiation fields at HFEF-N are known to have ranged from 

immeasurable levels to over 1 x 106 mrem/hr from irradiated fuel assemblies. However, the irradiated 

fuel assemblies emitting those radiation levels were always heavily shielded inside the HFEF-N’s hot 

cells or inside heavily shielded casks. 

FASB: Based on a review of the available ANL-W records, the photon radiation fields at the FASB 

were predominately from unirradiated uranium. Photon energies for uranium facilities are 

predominately in the range of 30-250 keV. Because the focus of data capture efforts was on 

contamination survey records versus radiation survey records, little is known about the actual photon 

radiation levels within the FASB. However, the photon radiation fields in the FASB would have 

ranged from immeasurable levels to those associated with large quantities of recycled highly-enriched 

uranium. 

ITF: Based on a review of the available ANL-W records, the photon radiation fields at the ITF were 

predominately from slightly-irradiated reactor fuels, activated materials, X-ray generating equipment, 

and un-irradiated uranium. Photon energies were predominately greater than 30 keV (ORAUT-
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TKBS-0007-6). Photon radiation fields at the ITF are known to have ranged from immeasurable 

levels up to 700 mrem/hr from slightly-irradiated fuel assemblies. 

TREAT: As is the case with any reactor, the potential existed for TREAT personnel to receive 

external exposures due to emissions from mixed fission and activation products during activities 

associated with reactor operation and maintenance. Reactor shielding permitted personnel access 

around and atop the reactor during steady-state operations at 100 kW. Access to the subpile room was 

controlled during steady-state operation and, prior to transient operations, the building was evacuated 

of all personnel (TREAT Manual, 1971). 

Radiation levels were monitored via a Jordan Electronics Model "RAMS II'' with an alarm and 

metering panel at the Control Building. Detectors were located in the subpile room, on the main floor, 

on the reactor-coolant exhaust filters, and on the exhaust stack. The monitor was set to alarm any time 

the main floor radiation level exceeded 7.5 mr/hr or the subpile room radiation level exceeded 50 

mr/hr (TREAT Manual, 1971). 

Photon exposure resulted from photon radiation produced during nuclear fission. In addition, many of 

the fission and activation products resulting from fuel irradiation decay by photon emission. The 

workplace photon energy spectra in the TREAT operational areas would have been dominated by 

high-energy photons from mixed fission and activation products. General gamma radiation levels at a 

distance of 10 feet from the reactor during operations at 100 kW were 5-8 mrem/hour (Freund, 1960). 

ZPPR: The possibility of external exposure existed at ZPPR as a result of working with reactor 

processes and fuel loading and unloading. Photon exposure resulted from photon radiation produced 

during nuclear fission. In addition, many of the fission and activation products resulting from fuel 

irradiation decay by photon emission. The workplace photon energy spectra in the ZPPR operational 

areas would have been dominated by high-energy photons from mixed fission and activation products. 

L&O Building: Photon exposure potential existed from the mixed fission and activation products from 

the reactor fuels analyzed within the L&O Building. The workplace photon energy spectra in the 

L&O Building would have been dominated by high-energy photons from mixed fission and activation 

products. Pa-234m is a decay product in the U-238 decay chain and emits a 2.29 MeV beta particle. 

A significant quantity of photons resulting from Bremsstrahlung radiation are produced and contribute 

photons of intermediate energy (30-250 keV). Although enriched uranium has significantly less 

in-growth of Pa-234m, U-235 and its decay products emit a 185.7 keV photon 57% of time and a 

143.8 keV photon 11% of the time. 

The most significant gamma activity would have been from Cs-137 decaying to Ba-137m and its 

subsequent gamma emission, as well as the decays of Eu-154, Ru-106, and Ce-144. To a lesser 

degree, gamma activity would also result from the decay of Rh-106 and Pr-144. 

In addition to the fission-product gamma rays, gamma rays from the activation of fuel cladding and 

structural materials (e.g., Mn-54, Co-58 and Co-60) would also be present, but would vary depending 

on the type of fuel under analysis at the time. The cells (caves) of the analytical junior cave facility 

were designed to safely handle analytical samples with a radioactivity level of 20 Ci of 1-MeV gamma 

radiation (EBR-II Fuel Cycle Story, 1987). 
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5.3.2.2 Beta

EBR-II: Based on a review of the available ANL-W records, the beta radiation fields at the EBR-II 

reactor were generated primarily by the operation of the nuclear reactor, irradiated reactor fuels, 

activated materials, releases of radioactive noble gases, and radioactive contamination from those 

materials. All beta radiation at the EBR-II reactor is attributed to greater than 15 keV electron 

radiation (ORAUT-TKBS-0007-6). 

FCF: Based on a review of the available ANL-W records, the beta radiation fields at the FCF were 

generated primarily by irradiated reactor fuels, activated materials, and radioactive contamination 

from those materials. All beta radiation at the FCF is attributed to greater than 15 keV electron 

radiation (ORAUT-TKBS-0007-6). 

HFEF-N: Based on a review of the available ANL-W records, the beta radiation fields at the HFEF-N 

were generated primarily by irradiated reactor fuels, activated materials, and radioactive 

contamination from those materials. All beta radiation at the HFEF-N is attributed to greater than 

15 keV electron radiation (ORAUT-TKBS-0007-6). 

FASB: Based on a review of the available ANL-W records, the beta radiation fields at the FASB were 

predominately from un-irradiated uranium. All beta radiation at the FASB is attributed to greater than 

15 keV electron radiation (ORAUT-TKBS-0007-6). 

ITF: Based on a review of the available ANL-W records, the beta radiation fields at the ITF were 

predominately from slightly-irradiated reactor fuels, activated materials, and un-irradiated uranium. 

All beta radiation at the ITF is attributed to greater than 15 keV electron radiation (ORAUT-TKBS0007-6).

TREAT and ZPPR: Reactor operations in the TREAT and ZPPR operational areas did not have pure 

beta external hazards, but beta exposures would have resulted from the mixed fission and activation 

products produced by the reactors. Whether beta radiation was considered an external hazard 

depended on the maximum energy of the beta emission for a given radionuclide, any shielding used, 

and the use of protective clothing. 

L&O Building: As with photon exposures, beta exposure potential resulted from the mixed fission and 

activation products from the reactor fuels analyzed within the L&O Building. Most fission products 

are initially rich in neutrons and undergo beta decay. In general, the thick shielding afforded by the 

cell walls would attenuate any beta particles emitted.
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5.3.2.3 Neutron 

Operating reactors would have been the primary source of neutron exposure during the period under 

evaluation. These reactors were experimental facilities for measuring shielding factors, reactor 

kinetics, and neutron irradiation. Thermal neutron radiation levels 10 feet from the TREAT reactor 

during operations at 100 kW were 50-1,500 n/cm2/sec (Freund, 1960).

Locations and periods of where neutron exposures are considered to be likely within the EBR-II 

Complex are identified in Table 6-11 of ORAUT-TKBS-0007-6. The EBR-II Complex areas listed in 

that table include TREAT (1958–1994), ZPPR (1969–1992), and the HFEF-N’s NRAD Facility 

(1977–2006) (ORAUT-TKBS-0007-6). More-than-incidental exposures to neutrons were not likely at 

the L&O Building, FASB and ITF because of the materials and activities being performed at those 

facilities. The following provides additional details for the EBR-II Complex facilities with 

potentially-significant sources of neutrons. 

EBR-II: Based on a review of the available ANL-W records, the neutron radiation fields at the EBR-II 

reactor were generated primarily by the operation of the reactor and irradiated reactor fuels. Neutron 

radiation fields at the EBR-II reactor are known to have ranged from immeasurable levels to over 

1 x 106 mrem/hr in the reactor core. In occupied areas of the reactor, neutron radiation fields were 

normally immeasurable (Neutron Surveys, 1967). 

FCF: Based on a review of the available ANL-W records, the neutron radiation fields at the FCF were 

generated primarily by irradiated reactor fuels, and a Sb-Be neutron radiography source (September 

1968 and later). However, these sources of neutrons were always in either heavily-shielded casks or 

the FCF hot cells. In occupied areas of the FCF, neutron radiation levels were negligible. 

HFEF-N: Based on a review of the available ANL-W records, the neutron radiation fields at HFEF-N 

were generated primarily by irradiated reactor fuels and the NRAD Facility’s reactor (October 1977 

and later). Neutron energies were generally less than 20 MeV (ORAUT-TKBS-0007-6). Neutron 

radiation fields at the NRAD Facility’s reactor are known to have ranged from immeasurable levels to 

over 1 x 106 mrem/hr in the reactor core. 

ZPPR: The operating reactor would have been the primary source of neutron exposure during the 

period under evaluation. These reactors were experimental facilities for measuring shielding factors, 

reactor kinetics, and neutron irradiation. Neutron and gamma dose rates were measured in ZPPR 

since start-up in 1969. In general, the neutron fields experienced by ZPPR personnel were relatively 

insignificant (less than 10 mrem/hr.) with the exception of those fields experienced between the 

reactor halves when the personnel shields were removed (Neutron Exposure Assignments, 1977). 
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5.4 Incidents at the ANL-W Site

5.4.1 Incidents at the EBR-I Complex

ATTRIBUTION: Section 5.4.1 was completed by Mike Mahathy, Oak Ridge Associated Universities, 

and Mitch Findley, MJW Corporation. All conclusions drawn from the data regarding the EBR-I 

Complex were peer-reviewed by the individuals listed on the cover page. The rationale for all 

conclusions in subsequent sections of this document devoted to the EBR-I Complex are explained in 

the associated text.

ANL-W facilities and activities involved experimental reactor design and development, closed-cycle 

irradiated-fuel processing, and analytical support. The ANL-W monitoring and analytical programs 

were designed to initiate an investigation of any potential internal intake, as indicated by off-normal 

workplace indicators such as personnel contamination or elevated air sampling. Although incidents 

involving exposures of regulatory significance were rare, the following incidents are provided as 

important events in ANL-W’s radiological history.

5.4.1.1 Excursion at BORAX-I (July 1954) 

A deliberate excursion to destroy the reactor was planned and carried out on July 22, 1954 to 

determine the inherent safety of the system under extreme conditions (ANL-W History, 2006; QA 

Assurance Files, 1991). After the test run of the reactor, one of the control rods became bowed in the 

subassembly, preventing the start of the excursion. Some of the staff were instructed to fix the rod so 

that the excursion could proceed. Two workers went to the reactor and, looking down, saw the 

unmistakable Cerenkov glow. One of the workers who was monitored for external dose reached into 

the pit to straighten the rod (Haroldsen, 2008).

Within a few milliseconds after the start of the excursion (following removal of the control rod from 

the reactor), a dark grey column of smoke and water was ejected from the reactor to a height of more 

than 80 feet (see Figure 5-48). Shortly thereafter, the sound of a medium-sharp detonation reached the 

control room about a half-mile away. Fuel plate fragments were scattered up to 300 feet, but no 

widespread dangerous dispersal was observed. Most of the heavy debris fell to the ground near the 

shield pit. The control rod drive mechanism fell on the side of the earth shield after having been 

thrown about 30 feet into the air. Fragments of fuel plates were thrown up to 200 feet from the reactor 

site. A photographer was sent to retrieve the film of the explosion; he then had to be decontaminated 

of fresh mixed fission products. Over the following days, workers were rotated into the reactor pit to 

recover gold foils so that they could determine the power the reactor attained before exploding. 

Wearing protective clothing and respirators, site staff retrieved materials and usable equipment from 

the reactor and surrounding area over the next two months (Haroldsen, 2008; Personal 

Communication, 2015b). Health Physics monitored clean-up and recovery operations (Griffiths, 

1954). Workers would pick up a shovel full of sludge and hold it to a detector to observe the count 

rate to determine the presence of a gold foil. 
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136 of 260 

Surveys of the total fission-product radioactivity of all the debris revealed that most of the fuel 

originally in the reactor could be accounted for within a 350 ft.-radius. No large fraction of the reactor 

core material left the site in the form of airborne material. More than fifty samples were analyzed for 

alpha activity from near the reactor through the site periphery. The samples included both filter paper 

samples of air particulates from the continuous air monitors permanently located around the ANL-W 

periphery, and gummed paper fallout plates from special stations established for the Borax 

experiment. In most cases, the activity of the air samples was statistically zero. Each sample 

represented from one to five days of continuous sampling. Health Physics analyzed three samples of 

contamination from Borax using a sodium iodide crystal gamma spectrometer to determine the 

energies of the contamination. They compared the spectrums of each to a spectrum of three-year-old 

fission products, and to a spectrum of six-day-old fission products from a test performed at the Pacific 

Proving Grounds in May 1954. 

Source: http://www.ne.anl.gov/About/reactors/lwr3.shtml

Figure 5-48: Destructive Test of BORAX-I 
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137 of 260 

The BORAX spectra closely compared to the spectrum from the Pacific Proving Grounds (Griffiths, 

1954). Figure 5-49 shows spectra from two of the contaminated BORAX-I items (left and center) and 

from the Pacific Proving Grounds (right). A modeled reconstruction of the excursion conducted by 

INL in 1991 provided estimates of fission products released by the excursion (QA Assurance Files, 

1991).

Samples counted within one day after termination of sampling had alpha activities ranging from 32 to 

200 dpm. The recounting of samples after two to three days showed the alpha activity to be the decay 

of radon and thoron daughters (Griffiths, 1954). A 1980 study stated that about twelve percent of the 

uranium in the core, or 479 g, was recovered from fuel fragments in the reactor pit and area 

surrounding BORAX (Smith, 1980). A modeled reconstruction of the excursion conducted by INL in 

1991 estimated that 4.7 g of uranium was released to the environment (QA Assurance Files, 1991). 

The balance of the uranium remained in the melted core (Smith, 1980). 
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Page 138 of 260 

Figure 5-49: Spectra Comparison: BORAX-I and Pacific Proving Grounds
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Page 139 of 260 

5.4.1.2 Suspected Brick Rupture in EBR-I Breeding Blanket (April 1955) 

In April 1955, transferrable contamination was discovered at the EBR-I reactor. A brick in the 

breeding blanket surrounding the reactor was suspected to have ruptured. Workers of concern were 

identified and sampled for beta and plutonium urinalysis. The available bioassay results did not 

indicate intakes of radioactive material occurred due to this incident (Bioassay Sheets, 1963, PDF pp. 

32, 54).

5.4.1.3 Partial Meltdown of EBR-I Core (November 1955) 

On November 29, 1955, the EBR-I core suffered a partial meltdown while performing a planned 

experiment for taking the reactor critical at a very low power level without coolant flow. The purpose 

of the experiment was to obtain transient temperature coefficients on fuel only. The meltdown 

occurred due to a misunderstanding by the reactor operator to use the slower control rods to shut down 

the reactor instead of the fast-acting safety rods. The two-second time difference was enough to 

permit the temperature of the reactor to overshoot to a point at which the alloying of uranium and steel 

and melting of uranium took place. Although the EBR-I reactor operated with an argon cover gas, 

fission products from the meltdown broke through and leaked into the reactor building. An order to 

vacate the reactor building was made. A large comprehensive special bioassay program followed. 

(Kittel, 1957; Zinn, Jun1956; Bioassay Sheets, 1963; Haroldsen, 2008). The bioassay results 

indicated there were no internal depositions of radioactive material as a result of the partial meltdown 

(Bioassay Sheets, 1963).

In March 1956, four months after the meltdown, work began on opening the reactor to examine the 

extent of damage. The time delay was essential to allow for an extended period of radioactive decay 

and to ease the procedure for lifting the core holder and fuel elements from the reactor. A concrete 

“hot” cave was constructed on the top of the reactor to provide shielding from the high radiation levels 

for the anticipated high external dose rates that were expected when the reactor assembly was lifted 

out of the reactor tank (Disassembly Compliance Report, 1956). The core was removed and placed in 

a custom-made transfer coffer that was sent to ANL in Illinois for examination. Air monitoring and 

contamination surveys were performed during the core removal (Air Sample Data Sheets, 1956).

5.4.1.4 Personnel Contamination at BORAX-IV (July 1957) 

On July 17, 1957, a [job category redacted] got a small amount of a radioactive waste water sample in 

[body part redacted]. Health Physics surveyed the [job category redacted] [body parts redacted] with 

no measureable activity found. Health Physics collected a bioassay sample that was reported as 

negative for activity (Safety Monthly Reports, 1957, PDF p. 80). 

5.4.1.5 Airborne Contamination at BORAX-IV (September 1957) 

A leak in the air ejector filter occurred in September 1957 during an extended run of the BORAX-IV 

reactor. Airborne contamination was detected by Health Physics. Additional air samples were 

collected and processed the following day; results indicated no activity above background. The filter 

on the air ejector was repaired (Safety Monthly Reports, 1957).
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5.4.1.6 Fuel Defect Test at BORAX-IV (March 1958) 

BORAX-IV was not operated at a power level from December 5, 1957 until February 19, 1958 when 

a power run was attempted. In the interim, ten new fuel elements were added to periphery of the core. 

After attempting to start the reactor on February 19, and again on February 25, radioactivity was 

measured in the reactor hotwell and turbine buildings at five times the maximum design activity 

(Bailey, 1958). A steam break also occurred in the turbine during the February attempt (Findings 

Report, 1958). Those start-up attempts were halted before maximum power was reached because 

ANL-W assumed that one of more fuel elements had rupture, but remained encapsulated by the cans. 

ANL-W planned controlled excursions for March 11-12 to determine the extent of the rupture and the 

limiting effects encountered while powering the reactor with defected fuel (Bailey, 1958). ANL-W 

Health Physics helped plan the test and provided monitoring during and after the test, which was 

concluded on March 12, 1958 (Status Summaries - Borax IV, 1958; Memo of Run, 1958; Defective 

Fuel Experiment, 1959). Dose rates were measured around the reactor and turbine buildings using 

Juno survey meters and neutron detectors. Health Physics also monitored reactor water, steam, and air 

ejector gases for radionuclides. Dose rates resulting primarily from N16[O16(n,p)N16] were measured 

ranging from 27 mr/hr at the main doors to the turbine and 30 mr/hr at the reactor buildings to 50,000 

mr/hr at the turbine air ejector exhaust filter (Borax-IV Test, 1959). Xe-138 and Kr-88 accounted for 

the major portion of radioactivity released from the fuel elements. The major contaminant released 

from the steam system was Cs-138 (Borax-IV Test, 1959; Findings Report, 1960). I-131 was detected 

in reactor water, steam, and condensate. Long-lived radionuclides that were identified in trace 

amounts in the turbine and condenser after operation were Ba-140, La-l40, Sr-90, Y-90, Sr-89, and 

Cs-l37 (Borax-IV Test, 1959). Alpha radiation was not detected in water, steam, or air. Only trace 

amounts of fission products without nongaseous precursors (e.g., Mo-99) were detected in the reactor 

water (Borax-IV Test, 1959, PDF p. 31). 

Personnel involved in the February and March releases were monitored for external radiation, and for 

fission product intake via urinalysis (Memo of Run, 1958; Defective Fuel Experiment, 1959). NIOSH 

has obtained those urinalysis results (Bioassay Sheets, 1963). 

5.4.1.7 Fire in EBR-I Rod Wash Room (September 1959) 

On September 17, 1959, an explosion in the rod washroom occurred when a pump and fan unit for the 

washroom exhaust caught fire. The fire spread to the wooden frame around the unit and part of the 

media of a filter used in the exhaust system. Air samplers were placed into service to determine if 

airborne activity escaped into the EBR-I basement area. Air samples taken during and after the event 

did not show significant airborne contamination. All bioassay of personnel involved were less than 

the limits of detection (Safety Monthly Report, 1959).

5.4.1.8 Contamination on Plutonium Plates at ZPR-III (March 1960) 

On March 14, 1960, ZPR-III staff were preparing for the initial use of plutonium fuel plates. These 

plates were clad in stainless steel. An inspection of the plates discovered “a small number” of plates 

with external contamination that was confirmed to be plutonium. While there were no intakes of 

radiological significance from the incident, it was the only time in the history of ZPR-III that evidence 

of plutonium contamination was found. Due to the recognized hazards of plutonium, a rigorous 
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program was established for surveying plutonium fuel plates, not just upon receipt, that sustained for 

ZPR-III’s operating history (Safety Monthly Report, 1960; Fish, 1967).

5.4.1.9 Alpha Contamination at AFSR (November 1963) 

On November 29, 1963, while removing irradiated U-235 foils from the reactor, a work table used for 

unloading the foils was contaminated to 200 dpm/100 cm2

 alpha. The table was subsequently 

decontaminated to less than 10 dpm/100cm2 alpha (Safety Monthly Reports, 1963).

5.4.1.10 Defective Fuel Test at BORAX-V (August 1964) 

Much like the fuel defect tests conducted at BORAX-IV, ANL-W prepared for power tests at 

BORAX-V using prefabricated defective fuel. Defective fuel elements were prepared at ANL to test 

power generation using superheated defective fuel. The canned defective elements replaced some of 

the normal fuel elements. The core was operated from the first week in August until August 14, 1964, 

producing 266 MW with the core containing partially-defective fuel. That final experiment in 

BORAX-V demonstrated that negligible contamination to turbo-generator equipment resulted from 

operating with an experimentally-defective fuel element in the superheated core (Borax Superheater, 

1964). 

ANL-W Health Physics monitored the August power generation. The major contaminants released 

from the steam system (as measured by air monitoring or smear analysis) were: Co-58, Kr-88, Sr-91, 

I-131, Xe-138, Ba-140, and La-140 (Progress Report, Aug1964). Personnel working at BORAX-V 

were routinely monitored by film badge and bioassay, including whole-body counting and urinalysis. 

5.4.1.11 Spontaneous Combustion of U-235 Foils at ZPR-III (November 1966) 

On November 23, 1966, enriched-uranium (93.19%) foils in paper envelopes caught on fire in the 

ZPR-III Special Materials Vault as they were being removed from a sealed storage container. The fire 

was quickly extinguished. Alpha contamination was found on the outer clothing and person of 

[number redacted] workers. The workers were decontaminated and sent to Central Facilities for an 

in-vivo count. The results of the counts were less than detection limits. In addition, [sample type 

redacted] samples were also requested of the [number redacted] workers. A gamma count of the 

[sample type redacted] samples indicated detectable activity but not significant intakes. Clean-up 

operations included smears and air samples taken in the vault, Loading Room, Control Room, 

Airlock, Assembly Room, and corridors. Decontamination efforts were taken until the floors were 

restored to < 5 dpm/100cm2 transferrable alpha contamination (Incident Report, 1966; Incident 

Summary, 1967).

5.4.1.12 Leaking Po-Be Source at ZPR-III (April 1967) 

On April 14, 1967, the facility [job title redacted] suspected a possible leaking Po-Be source during a 

source exchange at ZPR-III. External contamination of 1800 dpm/100 cm2 was discovered on a smear 

taken on the source. An air sampler in the Assembly Room (where the source was located) showed an 

“insignificant” amount of alpha activity. Analysis of the original smear indicated the contaminant to 

be Po-210. [Sample type redacted] and a [sample type redacted] were requested of the [job title 

redacted]. No activity was detected in either sample (Waste Discharges, 1958).
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5.4.2 Incidents at the EBR-II Complex

ATTRIBUTION: Section 5.4.2 was completed by Jason Davis, Oak Ridge Associated Universities, and 

Brian Gleckler, Dade Moeller, Inc. All conclusions drawn from the data regarding the EBR-II 

Complex were peer-reviewed by the individuals listed on the cover page. The rationale for all 

conclusions in subsequent sections of this document devoted to the EBR-II Complex are explained in 

the associated text.

ANL-W facilities and activities involved experimental reactor design and development, and irradiated 

fuel processing. ANL-W’s monitoring and analytical programs were designed to initiate an 

investigation of any potential internal intake indicated by off-normal workplace indicators, such as 

personnel contamination or positive air sampling. NIOSH found extensive personal monitoring data 

for mixed fission and activation products (beta/gamma); therefore, only external exposures of 

regulatory significance, significant alpha-contamination incidents, and criticality incidents are 

presented below. Several other incidents are presented to illustrate the effectiveness of radiological 

controls in preventing the spread of contamination and minimizing the chance of a significant intake 

of radioactive material.

5.4.2.1 Perchloric Acid Explosion in L&O Chemistry Lab 40 (July 1962) 

A perchloric acid explosion occurred in Chemistry Laboratory 40 on July 13, 1962. A glass beaker 

contained radioactive xenon and krypton gases absorbed on charcoal. The material was being 

dissolved in perchloric acid; the experiment was inadvertently allowed to go to dryness and an 

explosion occurred. Most of the debris was confined to the hood; however some glass fragments fell 

to the floor outside the hood. No activity was detected outside the hood; debris inside the hood was 

disposed of as dry active waste (Safety Monthly Report, Jul1962, PDF p. 143). 

5.4.2.2 L&O Building Broken Specimen (March 1963) 

On March 4, 1963, it was determined that some unirradiated, clad uranium-fissium alloy specimens 

had broken while being heated in an electric furnace. This triggered a survey of the entire Metallurgy 

Laboratory for both alpha and beta/gamma contamination using portable instruments and dry smears. 

No contamination was found. A special air sample indicated that no material was released into the air. 

The specimens in question were removed from the furnace, sealed in polyelthylene sleeves, and 

placed in dry active waste. A survey of the furnace after cool-down showed no contamination present 

on the interior surface (Safety Monthly Reports, 1963, PDF p. 130). 

5.4.2.3 Hydrogen Explosion in EBR-II/FCF Transfer Lock (Dec. 1965) 

On December 26, 1965, a hydrogen explosion occurred in the transfer lock between the EBR-II 

Reactor and FCF while the empty IBC was in the lock. This resulted in minor damage to the coffin 

and the lock. Specific requirements of the operating procedures limited the time an IBC containing a 

subassembly could be in the lock, but no such requirements were designated for an empty coffin. The 

empty IBC was placed in the transfer lock with the blower shut off. With the blower shut off, the 

battery charger was thus inadvertently placed on a high rate of charge, which caused hydrogen to be 

generated. It was surmised that, when the lock cover was actuated some hours later, a spark from a 

door limit switch ignited the hydrogen. The force of the resulting explosion blew the lid off the lock 

and bent the hinge; fortunately, no personnel were injured (Stevenson, 1987, PDF p. 72). 
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5.4.2.4 EBR-II Reactor Radioactive Sodium Fire (June 1967) 

A small radioactive sodium fire occurred in the primary sodium-sampling room. Thereafter, access to 

the sampling room was restricted until the Na-24 radioactivity had decayed. Some Na-24 was 

transported out of the sampling room by the ventilation system, and caused minor air and surface 

contamination in the reactor building (Progress Report, Jun1967, PDF pp. 47-48). 

5.4.2.5 FCF Contaminated Water Spill (June 1968) 

In the FCF Reactor Passageway, a spill of subassembly wash water overflowed from an inter-building 

coffin (IBC) when the gate valve at the top of the IBC was not closed properly. The contaminated 

water spread on the floor of the Reactor Passageway and left dry residues emitting as high as 

3E+04 dpm/100 cm2. The floor was mopped to remove the residues (Stevenson, 1987, p. 242).

5.4.2.6 ZPPR-AFSR Dropped Blanket Drawer (January 1975) 

On January 20, 1975, while spot-checking the loading for Assembly #5, a blanket drawer was 

dropped, striking eight fuel drawers that had been pulled out of the matrix about 3 inches. The cell 

was immediately evacuated via the work room. Radiological controls personnel were notified and the 

mound area placed in a state of isolation. Personnel surveys of the loading crew revealed no 

contamination and the crew was released from the work room. 

Re-entry was made by a team of three men, one of whom was from Radiation Safety. Each man wore 

a self-contained breathing apparatus. Smears were taken. Direct surveys were found to be unreliable 

because the depleted uranium from the blanket drawer had spilled out and broken, the pieces and 

powder scattering over the fuel drawers. There were no high readings on the alpha meter even under 

this condition, the highest being around 300 cpm. One fuel plate had visible damage and one smear 

was taken on this plate. The drawers were bagged, taken to the work room and placed in the hoods for 

containment. No high alpha activity could be detected on the smears with an alpha meter. 

Personnel surveys of the re-entry team revealed no contamination and they were released from the 

work room. The smears were counted and highest alpha activity noted was 150 dpm/100 cm2. Three 

smears with activity between 120 to 150 dpm/100 cm2 were selected for analysis. One of these was a 

smear taken on the visibly-damaged fuel plate. Two smears were sent to CF 690 for analysis, and one

was kept at ZPPR for analysis. Recovery operations were halted until analysis results were in. The 

analysis results were identical on all three smears and indicated the presence of depleted uranium with 

no evidence of Pu. 

Recovery began by removing the fuel from the drawers for inspection and contamination surveys. 

Smears samples were taken of each the 42 fuel plates. The highest activity recorded was 90 dpm/plate 

alpha. One more plate was found to be slightly dented. The two dented wafers were isolated in a 

sealed canister after they were decontaminated to <10 dpm/plate alpha. The rest of the fuel was 

isolated in sealed canisters and all canisters taken to the vault. Reflecting an overabundance of 

caution, this work was done with full-face respirators despite the work being performed under hoods. 

An air sample taken in the breathing zone decayed to 1.1 x 10-13 μCi/cc in 20 hours. 

Decontamination of the cell and work room hoods began and continued throughout the next day. 

Activity was reduced to less than 15 dpm/100 cm2

 alpha and less than 100 dpm/100 cm2 beta/gamma. 
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Air samples decayed to less than guidance levels for Pu after 24 hours of decay (Weekly Report, 

1975, PDF pp. 5-6).

6.0 Summary of Available Monitoring Data for the Class Evaluated 

by NIOSH

The following subsections provide an overview of the state of the available internal and external 

monitoring data for the ANL-W class under evaluation. 

6.1 Available ANL-W Internal Monitoring Data

ATTRIBUTION: Section 6.1 was completed by Mike Mahathy and Jason Davis, Oak Ridge Associated 

Universities, Brian Gleckler, Dade Moeller, Inc., and Mitch Findley, MJW Corporation. All 

conclusions drawn from the data regarding the specified facilities were peer-reviewed by the 

individuals listed on the cover page. The rationale for all conclusions in subsequent sections of this 

document devoted to the specified facilities are explained in the associated text. 

The Health and Safety Division of AEC’s Idaho Operations furnished a comprehensive personnel 

metering program for radiation monitoring of all Idaho site personnel. The Analysis Branch of the 

division initially provided routine urinalysis of all personnel with sampling frequency varying 

between quarterly and yearly, depending on the expected potential for uptake (Annual Report, 1959).

Chemical and radiochemical urinalysis were run at three-month intervals on all people who had 

frequent contact with radioactive material and at six-month intervals on those people who had only 

occasional contact. Analyses were also made annually on about fifty NRTS personnel who were not 

exposed to radiation to provide a statistical control group (Health Physics Services, 1952).

Site-wide routine samples were initially analyzed only for gross alpha and beta, with isotopic analysis 

performed only if activity was discovered in the sample, although capabilities existed for analyses of 

specific isotopes, such as Pu-239, Sr-90, Co-60, and I-131 in body excreta, water, and air samples. 

The Analysis Branch also made analyses for materials related to good industrial hygiene practices, 

such as beryllium, lead, mercury, arsenic, fluorine, or any other chemical that may have been required 

(Annual Report, 1959). 

Based on the loss of iodine due to volatilization during the wet-ashing process, as well as the small 

likelihood of a significant uptake of a pure beta-emitter without a correspondingly-large uptake of 

gamma-emitting fission products, the gross-beta procedure on urine samples was discontinued entirely 

in December 1960. However, in order to detect a slow build-up of the bone-seeking, beta-emitting 

strontium isotopes from low-level sources over long periods of time, specific strontium analyses were 

made for each individual who had worked in a reactor or chemical processing area for two years or 

longer (Murphy, 1966); these analyses were performed every two years and on termination of 

employment (Annual Report, 1961).

NIOSH has found only a small set of bioassay data for ANL-W workers for the period prior to 1958. 

During this evaluation, NIOSH learned that some monitoring data on early workers is stored at the 

INL records center and some at ANL in Chicago. In September 2015, NIOSH performed an 
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additional data capture at INL, partly to look for additional bioassay data for the period 1952 through 

1960. In October 2015, NIOSH also performed a review of available data at ANL for ANL-W 

workers. NIOSH submitted to both INL and ANL a list of 50 former workers believed to have 

worked at ANL-W prior to 1961, requesting available internal and external monitoring data. Partial 

data was returned for 48 of the 50 requested workers. Not all of the 50 workers were employed at 

ANL-W for the entire period 1952 through 1960. Table 6-1 lists the number of workers with bioassay 

records of any kind, by year, available to NIOSH for these 50 workers. Bioassay data reported for 

1952 represent baseline and incident monitoring. 

Table 6-1: Bioassay Results for 50 Selected ANL-W Workers

Year Subset of 50 Selected ANL-W Workers with Bioassay 

Results

1952 12

1953 None

1954 15

1955 24

1956 11

1957 5 

1958 23

1959 6 

1960 24

Both INL and ANL supplied bioassay monitoring data they each have available for the fifty former 

workers. However, data was not provided for some of the former workers, and for some of the years 

requested - varying by former worker. While there was some overlap of records supplied by INL and 

ANL, there is a clear trend in data reporting, by workers, as shown in Figure 6-1. Note that ANL 

reported bioassay data starting in 1952, but the number of workers for whom data was supplied trends 

downward. The trend for INL is reversed, which carries into the years after 1960. For 1952, INL 

provided data for only one former worker; ANL provided data for no former workers. For 1955, ANL 

provided bioassay data for two former workers; INL provided bioassay data for four former workers. 

Not each of the 50 former workers were employed at ANL-W each of the years from 1952 through 

1960. 
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Figure 6-1: Bioassay Monitoring Data Trend

Whole-body counting (WBC) started at the Idaho site in January 1961. The convenience, sensitivity, 

and reliability of WBC made it the choice of personnel monitors for internal radioactivity from 

gamma-emitting nuclides. Consequently, urinalysis as a routine monitoring technique was gradually 

phased out by the laboratory. From that time, urine was analyzed only for iodine, uranium, and 

strontium or for elimination studies on specific isotopes (Annual Report, 1969). Early in the site’s 

history, job duties and their corresponding exposure potentials were used to place employees into the 

following routine WBC frequencies: quarterly, yearly, or no count needed. Special analyses were 

requested by the relevant Health Physics sections when internal exposure was suspected due to events 

occurring in a plant; terminations that required a physical examination (as determined by the relevant 

Health & Safety sections) also required a whole-body count (see Figure 6-2) (Procedures, 1963).

A portable, unshielded whole-body counter was developed and constructed in 1963. A medical van 

containing a whole-body counter was designed and purchased in 1965. By 1969, WBC equipment 

available at the Analytical Chemistry Branch Laboratory included rotational counting, helical 

scanning, and a detector for determining Sr-90 in the skeleton by counting the Bremsstrahlung 

emission from the skull (Annual Report, 1969).
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Source: Bioassay, 1956-1984; Bioassay, 1956-1967; Bioassay, 1958-1986

Figure 6-2: Total Internal Monitoring Performed at ANL-W (1956-1979)

In May 1967, the Health Services Laboratory was organized from six branches of the Idaho 

Operations' Health and Safety Division. Its primary function was to develop a technical staff in 

occupational medicine, health physics, radiological dosimetry, analytical chemistry, and 

instrumentation to support the Idaho site’s occupational health programs (Annual Report, 1969). The 

Analytical Chemistry Branch had responsibility for estimating the dose received from internallydeposited radionuclides at the site until 1971. Sometime during 1971, the responsibility was 

transferred to a health physicist in the Environmental Sciences Branch (Methods, 1979).

ANL-W recognized the potential for plutonium uptake when handling plutonium-fuel-bearing 

assemblies and established a bioassay program to monitor those individuals known to be working with 

these materials. The program required that both a urine sample and a fecal sample be submitted 

annually for analysis. A total of 68 employees were included in this sampling plan in 1975: 32 from 

HFEF, 7 from TREAT, 21 from Radiation Safety, and 8 from ZPPR (Bioassay Instructions, 1975).

Details regarding the various analyses used and the associated minimum detectable activities are 

presented in the Technical Basis Document for INL - ANL-W - Occupational Internal Dose (ORAUTTKBS-0007-5).
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6.2 Available ANL-W External Monitoring Data

ATTRIBUTION: Section 6.2 was completed by Mike Mahathy and Jason Davis, Oak Ridge Associated 

Universities, Brian Gleckler, Dade Moeller, Inc., and Mitch Findley, MJW Corporation. All 

conclusions drawn from the data regarding the specified facilities were peer-reviewed by the 

individuals listed on the cover page. The rationale for all conclusions in subsequent sections of this 

document devoted to the specified facilities are explained in the associated text. 

Personal dosimeter data (i.e., film badge and/or TLD) are the primary data type used to reconstruct 

external doses for ANL-W workers. Those data are specifically used to reconstruct a worker’s 

measured and missed external doses. For the period under evaluation, the ANL-W site monitored all 

workers for external photon and beta (electron) doses via their own dosimeters when inside the major 

radiological operating areas (Cipperley, 1958; Cipperley, 1968; ORAUT-TKBS-0007-6). 

NIOSH does not possess a complete set of the ANL-W’s dosimeter data; however, the DOE does have 

a complete set for the years 1958 and beyond. With two exceptions, the DOE provides the external 

dosimetry records for each energy employee who files an EEOICPA claim: (1) the dosimetry records 

for ANL-W workers are sometimes missing or incomplete for the years prior to 1958; and (2) external 

dosimetry records are currently not retrievable for certain workers who were only issued temporary 

dosimeters. Readily-retrievable permanent records were not created for some ANL-W workers 

(mostly visitors and subcontractors) who met the following criteria: (a) only had temporary dosimeters 

assigned; (b) only had dosimeter results below the dosimeter limits of detection (i.e., no positive 

external doses reported); and (c) did not have any bioassay records. The records of those temporary 

dosimeter results are available at the site; however, they have not been indexed by the DOE and are 

not currently in a readily-retrievable format. NIOSH is currently in the process of capturing the 

temporary dosimeter records and putting them in a retrievable format for the ANL-W worker

EEOICPA claims.

A summary of the dosimetry data is provided later in Figures 6-4 and 6-5. 

NIOSH submitted to both INL and ANL a list of fifty former workers believed to have worked at 

ANL-W prior to 1961, requesting available internal and external monitoring data. Not all of the fifty 

workers were employed at ANL-W for the entire period 1952 through 1960. Both INL and ANL 

supplied external monitoring data they each have available for the fifty former workers. However,

data was not provided for some of the former workers, and for some of the years requested - varying 

by former worker. While there was some overlap of records supplied by INL and ANL there is a clear 

trend in reporting of data, by workers, as shown in Figure 6-3. Note that ANL reported external 

monitoring data for significantly more workers than did INL through 1955, after which ANL records 

drop off dramatically. For INL, the number of workers with reported external data steadily climbed 

from 1952 until 1960. 
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Figure 6-3: External Monitoring Data Trend

As early as 1955, ANL’s entire Idaho Division was considered an active area, making the wearing of 

personal monitoring devices by all employees mandatory. The badges worn at this time contained 

sensitive film for measuring fast neutrons, in addition to the usual beta and gamma films (Personnel 

Exposure Reports, 1962, PDF p. 43). 

In 1950, the Division of Medical Services was established and given the task of handling personnel 

monitoring for employees and contractors at the entire Idaho site. The Health Physics Division was 

established on June 30, 1951. This was the same year that EBR-I went "hot." The Personnel 

Metering Branch was established as one of seven branches of the AEC’s Idaho Operations Office 

Health and Safety Division. The Personnel Metering Branch had the responsibility for determining 

external dose. The Branch furnished personnel metering services to 21 prime service contractors 

scattered throughout 18 separated areas within the Idaho site (Cipperley, 1959).

In the early days, dose was determined by film only encased in a film badge. Later, other devices 

were employed, such as pocket chambers (Pencils), finger rings, and wrist badges. These devices 

were used to detect all forms of radiation except alpha (i.e., beta, gamma, and neutrons). The 

Personnel Metering Branch maintained complete records was on all persons at the Idaho site. These 

records were cataloged by assigned badge number and were permanently filed according to date and 

alphabetically by surname. These records were summarized quarterly and statistically evaluated for 

each person (Puphal, 1996).
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There were three types of film used in the badges serviced by the Personnel Metering Branch:

• The DuPont Beta-gamma Type 552 film packet was used in most of the area badges. Each film 

packet contained a piece of sensitive film and a piece of insensitive film along with a 1 mm 

cadmium filter. The sensitive film (502) had a threshold level of approximately 30 mR and 

became saturated at about 5 R. The insensitive film (510) had a threshold of approximately 150 

mR and became saturated at about 30 R. If the sensitive film became saturated, the insensitive 

film was read (Wilson, 1961). 

• The DuPont Type 558 film packet was sometimes referred to as a “Hanford-type” dosimeter. The 

packet’s sensitive film (508) had a range of 10 mR to 10 R and was insensitive from 10 R to 

approximately 800 R. The insensitive film (1290) had a range of 10 R to 1,000 R (Wilson, 1961).

• Kodak Personal Neutron Monitoring film, Type A, was used in the badges when neutron coverage 

was requested, with one piece of neutron film placed in each packet. This film was sensitive to 

neutrons with energies greater than 0.5 MeV; the minimum detectable level was approximately 14 

mrem. This film was worn in the badge with either of the types of beta-gamma film.

Over time, it became evident that better methods of exposure determination and more efficient record 

and reporting procedures were required. In March 1958, the Branch began the use of a modified 

version of the Hanford-type plastic film badge containing filters of 0.0393” cadmium, 0.005” silver, 

and 0.191” aluminum, providing absorber thicknesses of 950 mg/cm2

, 203 mg/cm2, and 175 mg/cm2

respectively (Cipperley, 1959).

Wrist badges were identical to the regular film badges except that the clips were replaced with

expandable wrist bands similar to those found on many wrist watches. 

Temporary badges were used primarily for visitors. A film packet was placed under a cadmium shield 

and secured by Scotch tape. The shield had a number perforated in it for identification when the film 

was exposed to X-rays. The cadmium shield covered only half of the film to show a difference 

between beta and gamma exposures. Any visitor, or any person using a temporary badge, was 

required to add his/her full name, "S" number (if an employee), business address and company 

represented. The badge was held by a clip like regular badges.

Finger rings were used in situations primarily involving radiation exposure to the hands. Originally, 

an aluminum ring with a silver filter was used, but this was replaced by a plastic ring with a cadmium 

filter. The rings were made up of three parts. The ring base and the finger band, a small filter, and the 

ring top. The film was placed in the ring base, covered by the filter, and the ring top was screwed on. 

Type 552 film was used in all rings (Puphal, 1996).

Film badge and finger ring services were provided to the EBR area on a weekly basis by the Personnel 

Metering Branch. Personnel exposure records were maintained by EBR staff (Personnel Metering, 

1958). During the early years, monitoring summary data contained only the total number of badges 

issued during a monitoring period (see Figure 6-4). In December 1958, records began to include a 

tally of how many persons participated in the monitoring program (see Figure 6-5). By January 1967, 

the total number of badges issued was dropped from the summary reports entirely; only the number of 

individuals monitored was provided. Also in 1967, the site began transitioning from weekly and 

monthly film dosimetry exchanges to quarterly exchanges of thermoluminescent dosimeters (TLDs), 
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resulting in fewer individuals being recorded on each month’s badge exchange listing (Annual Report, 

1959; Annual Report, 1960; Annual Report, 1961; Annual Report, 1962; Exposure Resumes, 1966; 

Exposure Resumes, 1970; Exposure Records, 1972; Exposure Records, 1973; Exposure Records, 

1974; Personnel Metering Tables, 1962; Personnel Metering Tables, 1963; Personnel Metering 

Tables, 1964; ZPPR Dosimetry Policy, 1971). Although ZPPR became operational in 1967, badges 

from that site were included in the EBR-II tally until November 1971, when they began to be reported 

separately (ZPPR Dosimetry Policy, 1971, PDF p. 8). 

Figure 6-4: ANL-W Badges Issued by Area (1952-1966)

NOTE: Figure 6-4 was developed using Personnel Metering Monthly Reports; NIOSH has not 

secured access to all of these reports. Gaps in the data display (e.g., early 1964) do not 

necessarily indicate a lapse in monitoring; rather, such gaps indicate a time period for 

which NIOSH does not have access to the monthly reports. 
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Figure 6-5: Number of Persons Monitored at ANL-W by Area (1958-1974)

NOTE: Figure 6-5 was developed using Personnel Metering Monthly Reports; NIOSH has not 

secured access to all of these reports. Gaps in the data display (e.g., early 1964) do not 

necessarily indicate a lapse in monitoring; rather, such gaps indicate a time period for 

which NIOSH does not have access to the monthly reports. 
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6.3 Available ANL-W Radiological Monitoring Data

Consideration of personnel exposures to radioactive material commenced before the start of 

radiological operations at ANL-W. The AEC-IDO took advantage of the practices and experiences of 

other AEC facilities established earlier, such as Oak Ridge National Laboratory and Hanford, to 

establish their health physics program. AEC-IDO policy regarding health physics responsibilities and 

services at INL was established in 1952 creating centralized services (Johnston, 1952), which included 

ANL-W even though ANL-W was under the AEC-Chicago Operations Office. ANL-W used air 

sampling and radiological contamination monitoring programs as qualitative indicators of internal 

exposures, with routine bioassay programs serving as verification programs that internal exposures 

had not occurred. 

The contamination control limits for the detection and control of released activity beyond the control 

boundaries related to instrumentation capabilities and the basic philosophy of acceptance of detectable 

contamination. The contamination control limits for on-plant surfaces, and particularly personnel, 

were always close to the MDA, so that any detectable contamination was a signal for follow-up 

evaluations and actions. 

The monitoring of radioactivity in the air in occupied areas was a fundamental component of the 

internal exposure control program. Beta/gamma CAMs were used from the beginning of all facility 

and program operations in routinely-occupied areas. The CAM systems provided real-time airactivity evaluations while fixed air samplers provided retrospective data and an average air 

concentration in an area or building. 

In 1958, an informal study and information program was initiated within the Idaho Division Health 

Physics section. This resulted from a series of requests from ANL-Idaho staff for data and 

information on the detection and monitoring of plutonium aerosols under conditions of reactor 

operations. The material assembled in the course of the program was originally written in the form of 

a series of information memoranda circulated within the Idaho Division. These memoranda were

revised, augmented, and updated to form a single report in 1963 that discusses the monitoring and 

sampling applications in the vicinity of operating nuclear reactors (Stoddart, 1963). The report

describes the technical bases for the plutonium aerosol monitoring program within the ANL-Idaho 

facilities, as summarized in the following paragraphs.

Continuous monitoring and sampling was provided in all areas where plutonium was present. 

Minimum coverage was considered to be a continuous "self-monitoring" sampler (either a 

conventional alpha air monitor or a ratio detection monitor), supplemented by one or more 

conventional air samplers. Monitoring units were to be inspected daily for proper operation. All 

monitors were equipped with chart recorders and all charts were kept as permanent records. Filters 

were removed from monitoring units and routinely counted as a double check on the monitor's 

detection and recording system (Stoddart, 1963). 

The conventional air-sampling units used were continuously-operating devices sampling at relatively

low flow rates. Typical units sampled room air at 2 cfm on 2-in.-diameter HV-70 or Millipore filters. 

Samples were removed daily, Monday through Friday, and counted for alpha and beta-gamma 

activities (Stoddart, 1963). 
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In addition to continuous area monitoring, all operations involving the handling, transfer, 

examination, or inventory of plutonium were monitored for release of plutonium aerosols by moving 

one of the continuous monitoring devices to the vicinity of the operation. At other times, specialpurpose monitors and samplers were used. Minimum coverage was considered to be provided by one 

continuous monitor supplemented by one or more conventional air samplers or special-purpose 

monitors (Stoddart, 1963). 

The air samplers used for the purposes noted above would have been either low-flow-rate devices (as 

described above) or high-volume types. The latter were preferentially selected when a specific 

operation was of short duration because adequate volumes could be sampled in a brief period 

(Stoddart, 1963). 

The low-flow-rate samplers used carbon-vane vacuum pumps and normally operated at flow rates of 1 

to 2 cfm. The high-volume or high-flow-rate samplers were high-speed metal-vane or rotary positivedisplacement devices. Sampling flow rates were dependent on the physical size of the filter, the type 

of filter, and the power rating of the unit. Depending on the unit and filter used, flow rates would 

have been on the order of 8 to 40 cfm (Stoddart, 1963). 

One example of a special-purpose monitor is an experimental “sniffer.” This unit, built in the Idaho 

Division Health Physics Section's instrument shop, consisted of a portable alpha survey meter coupled 

with an air-sampling device to provide a portable unit capable of measuring alpha particles 

simultaneously with deposition on the paper filter. Air was drawn through the filter at a rate of 10 to 

30 cfm. The unit was provided with a long “snout” to enable it to probe cavities, openings, and 

inaccessible areas (Stoddart, 1963). 

By Idaho Division Health Physics Section policy, sample media from conventional air samplers, 

continuous monitors, and impaction samplers are routinely counted for both alpha and beta-gamma 

activities. Discriminating counters were used to count for gross alpha and for gross beta-gamma. 

Alpha counters were calibrated against plutonium standards. Beta-gamma counters were calibrated 

against Radium D, E, F (combined) standards. The standards used were secondary standards, 

prepared and calibrated by the background counting group of Argonne's Industrial Hygiene and Safety 

Division. All counters in regular use were calibrated daily for yield or geometry and for residual 

background (Stoddart, 1963). 

Long-lived residual alpha activity (i.e., activity remaining after seven-day decay) was assumed to be 

plutonium unless positive isotopic or chemical identification was made. Operating procedures from 

the 1960s required isotopic identification on all samples having long-lived residual gross alpha 

activity greater than 50% of MPC(40) for plutonium after seven or more days of decay time. 

Identification was made by pulse-height analysis techniques (Stoddart, 1963). 
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6.3.1 Available ANL-W Radiological Monitoring Data for EBR-I Complex

The radiological monitoring for EBR-I, ZPR-III, AFSR, and BORAX I-V was provided by a central 

Health Physics group located in the EBR-I Complex. The radiological monitoring practices and 

policies employed would be applicable to all of these facilities. 

6.3.1.1 Available Radiological Monitoring Data for EBR-I 

ATTRIBUTION: Section 6.3.1.1 was completed by Mitch Findley, MJW Corporation. All conclusions 

drawn from the data regarding EBR-I were peer-reviewed by the individuals listed on the cover page. 

The rationale for all conclusions in subsequent sections of this document devoted to EBR-I are 

explained in the associated text.

Radiological monitoring in the form of routine survey, smear, and air monitoring was performed at 

EBR-I from the inception of operations. However, NIOSH found limited quantities and types of 

monitoring performed during the evaluation period. Many of the records found were located in the 

personal files of the EBR-I Complex Radiation Protection Manager; although the records were 

incomplete, they included data from important radiological events. Air monitoring data prior to 1955 

were not found. The amount of air monitoring data from 1955 into the early 1960s is limited, after 

which it was determined to be much more complete and comprehensive. Air samples were counted 

for beta/gamma and alpha contamination. Contamination and radiation surveys prior to 1959 were 

limited, but surveys after 1960 are available. NIOSH collected a sampling of survey, smear, and air 

monitoring data for EBR-I with an emphasis on collecting as much alpha monitoring data as possible. 

Final comprehensive surveys of the EBR-I facility were performed after decontamination and 

decommissioning operations were completed in 1975 in preparation for turnover of the facility to the 

National Park Service (Keating, 1975). 

Figure 6-6 shows an example of an air sampling data sheet for EBR-I recorded in 1955. Figure 6-7 

shows an example of an alpha contamination smear sample data sheet for EBR-I recorded in 1961. 

These are just examples of the hundreds of air sampling and contamination smear surveys that were 

found for EBR-I.
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Source: Sample Data Sheets, 1955

Figure 6-6: Example of Air Sampling Data Sheet at EBR-I in 1955
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Source: Sample Data Sheets, 1961

Figure 6-7: Example of Alpha Contamination Smear Survey at EBR-I in 1961

6.3.1.2 Available Radiological Monitoring Data for ZPR-III

Since the beginning of operations in 1955, radiological monitoring has been performed at ZPR-III. 

NIOSH was able to find examples of air monitoring from its initial years of operations (Air Sample 

Data Sheets, 1956; Air Sample Data Sheets, 1958). More comprehensive air monitoring records were 

found for later years. Radiation and contamination survey programs were employed from inception 

due to the handling of uranium fuel plates. This program became more rigorous with the introduction 

of plutonium fuel plates and the recognized potential for widespread uncontrolled contamination if a 

fuel plate rupture were to occur. The completion of a radiological monitoring “check sheet” was 

required each time plutonium fuel plates were handled. Figure 6-8 shows an example check sheet. 

One of the additional monitoring prerequisites was direct determination of surface alpha activity on all 

plutonium plates at routine intervals as well as before and after the handling of each plate. This 

procedure was called “piece counting.” Figure 6-9 shows an example piece-counting sheet. NIOSH 

was able to collect many examples of “check sheets” and the data associated with the checklists. 
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Source: Survey Reports, Jan1965

Figure 6-8: Example of Check Sheet Used for Plutonium Loading
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Source: Survey Reports, Jul1963

Figure 6-9: Example of “Piece Counting” of Plutonium Fuel Plates

6.3.1.3 Available Radiological Monitoring Data for BORAX

ATTRIBUTION: Section 6.3.1.3 was completed by Mike Mahathy, Oak Ridge Associated Universities. 

All conclusions drawn from the data regarding BORAX were peer-reviewed by the individuals listed 

on the cover page. The rationale for all conclusions in subsequent sections of this document devoted 

to the BORAX are explained in the associated text. 

From the onset of radiological operations through the end of the period under evaluation, health 

physics groups affiliated with ANL-W conducted air sampling and surveys for dose rate and 

contamination at the BORAX reactors. However, the degree and amount of radiological monitoring 
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data available to NIOSH is very limited for BORAX-I through Borax-IV. NIOSH has collected a 

sampling of routine survey, smear, and air monitoring data for BORAX-V through the end of reactor 

operation and dismantling. Radiation surveys included measurements of beta/gamma and neutron 

dose rates. Smears were collected for beta/gamma and alpha contamination while air monitoring 

filters were counted for fission products and alpha releases. Figure 6-10 shows an example of a 

radiation survey (Survey Reports, Jun-Aug1963) from the hundreds of pages of radiation surveys 

available to NIOSH for BORAX-V. 

Source Survey Reports, Jun-Aug1963

Figure 6-10: Example of Radiation Survey of BORAX-V Reactor Main Floor
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Figure 6-11 shows a portion of another example routine survey of the reactor main floor with results 

and mapped locations (Survey Reports, Jun-Aug1963).

Source: Survey Reports, Jun-Aug1963

Figure 6-11: Example of Smear Survey of BORAX Reactor Main Floor
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Figure 6-12 shows an example of air monitoring results for the BORAX-V turbine building (Survey 

Reports, Jun-Aug1963) from routine air monitoring data available for the reactor over its operational 

life. Note that the filter was recounted for alpha contamination to account for short-lived alpha 

products. 

Source: Survey Reports, Jun-Aug1963

Figure 6-12: Example of Air Monitoring Results for BORAX-V Turbine Building

The report in Figure 6-12 is indicative of all air monitoring reports for the EBR-1 Complex. Reports 

listed the location of each sample, total sampling time, air sampler flow rate, and other data needed to 

convert counts per minute to dpm/cm3

, date and times of the initial counts, and results on the initial 

counts. Each report also indicates if sampling was routine or performed to monitor for suspected 

airborne activity from loss of containment. In general, air sample data reports would indicate the 

suspected alpha radionuclide, although for BORAX-V it was always uranium. Samples were 

recounted when results were near or greater than ten percent maximum permissible concentration 

(MPC) to account for decay of radon. The alpha sample from Figure 6-12 was counted three times.
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6.3.1.4 Available Radiological Monitoring Data for AFSR 

ATTRIBUTION: Section 6.3.1.4 was completed by Mike Mahathy, Oak Ridge Associated Universities. 

All conclusions drawn from the data regarding the AFSR were peer-reviewed by the individuals listed 

on the cover page. The rationale for all conclusions in subsequent sections of this document devoted 

to the AFSR are explained in the associated text. 

ANL-W Health Physics conducted surveys for dose rate and contamination at AFSR from the onset of 

radiological operations in 1959 (at the EBR-I Complex) and after AFSR’s move to the EBR-II 

Complex in 1970. NIOSH has collected a representative routine survey and smear for AFSR as well 

as a vent fission product analysis performed in 1959. The early air monitoring demonstrated air 

activity involved short-lived radionuclides at well below ten percent MPC (Air Sample Data Sheet, 

1959). ANL-W Health Physics installed a continuous air monitor in AFSR in October 1959 (Air 

Sample Data Sheets, 1965, PDF p. 173). Its continued use is documented in September 1960 (Safety 

Monthly Report, 1960, PDF p. 73) and March 1961 with an average 36 dpm/m3 beta/gamma reported 

(Safety Monthly Report, 1961, PDF p. 217). However, NIOSH has not identified routine air 

monitoring data beyond 1959. 

Smears were collected to assess possible beta/gamma and alpha contamination. Figure 6-13 shows an 

example of a radiation survey that reports smears by AFSR location (Survey Reports, Jan-Feb1963). 

Records of additional surveys are available. Radiation surveys included measurements of beta/gamma 

and neutron dose rates. 
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Source: Survey Reports, Jan-Feb1963

Figure 6-13: Example of AFSR Radiation Survey with Smears
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The potential for neutron exposure was the primary concern of Health Physics at AFSR. ANL-W 

Health Physics monitored irradiations and reactor power tests for thermal and fast neutrons. Neutrons 

were measured by instruments that included the Rudolph Meter (a proton-recoil proportional detector 

for fast neutrons in the 0.2 to 10 MeV energy range) and the Fairport detector (a proton-recoil 

proportional detector with energy range 0.1 to 15 Mev). Neutron flux was reported with some neutron 

surveys (Air Sample Data Sheet, 1963, PDF p. 2; Neutron Survey Sheet, 1966, PDF p. 68). Figure 

6-14 shows an example of a neutron survey report. Additional reports are available.

Source: Neutron Survey Sheet, 1966, PDF p. 68

Figure 6-14: Example of ASFR Neutron Survey
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6.3.2 Available ANL-W Radiological Monitoring Data for EBR-II Complex

ATTRIBUTION: Section 6.3.2 was completed by Jason Davis, Oak Ridge Associated Universities, and 

Brian Gleckler, Dade Moeller, Inc. All conclusions drawn from the data regarding the EBR-II 

Complex were peer-reviewed by the individuals listed on the cover page. The rationale for all 

conclusions in subsequent sections of this document devoted to the EBR-II Complex are explained in 

the associated text.

The radiological monitoring at the EBR-II Complex was provided by a central Health Physics Group 

so that the radiological monitoring practices and policies would be consistently applied to all of the 

facilities. The Health Physics Group had satellite offices in several of the major facilities in the 

EBR-II Complex. The satellite offices typically served more than one facility and health physics 

technicians often rotated through the various satellite offices. As indicated by a former Radiological 

Controls manager, this rotation was intentionally scheduled to allow technicians to develop a 

familiarity with each of the ANL-W facilities (Personal Communication, 2015q).

From the onset of radiological operations through the end of the period under evaluation, ANL-W 

Health Physics performed routine and non-routine dose rate surveys, contamination surveys, and air 

monitoring at the radiological facilities within the EBR-II Complex. NIOSH has collected numerous 

samples of the available dose rate survey, contamination survey, and air monitoring data for the 

EBR-II Complex facilities for each year through the end of the period under evaluation. In areas 

where alpha contamination levels were potentially significant relative to the beta-gamma 

contamination levels, contamination smear samples were typically counted for both gross beta/gamma 

and gross alpha radioactivity. Contamination smear samples collected from areas where alpha 

contamination levels were not significant relative to the beta-gamma contamination levels were 

typically only counted for gross beta/gamma radioactivity. Air monitoring filters were counted for 

gross beta/gamma and gross alpha radioactivity, and sometimes received a second or third decay 

count for gross alpha radioactivity. 

6.3.2.1 Available Radiological Monitoring Data for TREAT 

Airborne activity in the TREAT Reactor Building was monitored with a cart-mounted continuoussampling unit with an integral recorder and an alarm that automatically reset. The detector’s range 

was 50-50,000 counts/min. A remote alarm was installed in the Control Building.

A criticality monitor was located at the north end of the reactor building. Normally, the monitor's 

green light would be on, but a radiation level of approximately l R would activate the alarm. This 

would cause the red light and the horn to oscillate on and off until the radiation level fell below the 

trip point. Malfunction of the monitor would also cause the alarm light and horn to be activated. Two 

Nuclear Accident Dosimeters (NAD) were also located in the reactor building. 

Figure 6-15 shows an example of a radiation survey with smears performed at the TREAT reactor 

room. Figure 6-16 shows a TREAT radiation survey report that corresponds to the survey displayed 

in Figure 6-15. 
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Source: Survey Reports, 1972

Figure 6-15: Example of TREAT Reactor Room Radiation Survey with Smears
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 Source: Survey Reports, 1972

Figure 6-16: TREAT Radiation Survey Report Corresponding to Figure 6-15 
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6.3.2.2 Available Radiological Monitoring Data for EBR-II Reactor

Figure 6-17 shows an example of a survey report for the EBR-II Reactor recorded in 1969. 

Figure 6-18 shows an example of a contamination smear sample data sheet for the EBR-II Reactor 

recorded in 1969. Figure 6-19 shows an example of an air sample data sheet for the EBR-II Reactor 

recorded in 1969. 

Source: Survey Reports, Aug-Dec1969

Figure 6-17: Example of EBR-II Reactor Survey Report from 1969
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Source: Survey Reports, Aug-Dec1969

Figure 6-18: Example of EBR-II Reactor Smear Sample Data Sheet from 1969
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Source: Survey Reports, Aug-Dec1969

Figure 6-19: Example of EBR-II Reactor Air Sample Data Sheet from 1969 
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6.3.2.3 Available Radiological Monitoring Data for FCF

Figure 6-20 shows an example of a survey report for the FCF main floor recorded in 1965. 

Figure 6-21 shows an example of an air sample data sheet for the FCF recorded in 1965. Figure 6-22 

shows an example of a contamination smear sample data sheet for the FCF recorded in 1965. 

Source: FCF Survey Reports, 1965, PDF p. 32

Figure 6-20: Example of FCF Survey Report from 1965 
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Source: FCF Survey Reports, 1965, PDF p. 98

Figure 6-21: Example of FCF Air Sample Data Sheet from 1965 
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Source: FCF Survey Reports, 1965, PDF p. 33

Figure 6-22: Example of FCF Smear Sample Data Sheet from 1965 
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6.3.2.4 Available Radiological Monitoring Data for FASB

Figure 6-23 shows an example of a survey report for the FASB recorded in 1978. Figure 6-24 shows 

an example of an air sample data sheet for the FASB recorded in 1978. Figure 6-25 shows an 

example of a contamination smear sample data sheet for the FASB recorded in 1973. 

Source: Survey Reports, 1978

Figure 6-23: Example of FASB Survey Report from 1978
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Source: Survey Reports, 1978

Figure 6-24: Example of FASB Air Sample Data Sheet from 1978
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Source: Survey Reports, 1973

Figure 6-25: Example of FASB Smear Sample Data Sheet from 1973
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6.3.2.5 Available Radiological Monitoring Data for ITF

Figure 6-26 shows an example of a survey report for the ITF recorded in 1969. Figure 6-27 shows an 

example of an air sample data sheet for the ITF recorded in 1969. Figure 6-28 shows an example of a 

contamination smear sample data sheet for the ITF recorded in 1969. 

Source: ITF Survey Data, 1967-1969

Figure 6-26: Example of ITF Survey Report from 1969 
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Source: Survey Reports, Feb1969

Figure 6-27: Example of ITF Air Sample Data Sheet from 1969
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Source: ITF Survey Data, 1967-1969

Figure 6-28: Example of ITF Smear Sample Data Sheet from 1969 
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6.3.2.6 Available Radiological Monitoring Data for HFEF-N 

Figure 6-29 shows an example of a survey report for the HFEF-N recorded in 1977. Figure 6-30 

shows an example of an air sample data sheet for the HFEF-N recorded in 1976. Figure 6-31 shows 

an example of a contamination smear sample data sheet for the HFEF-N recorded in 1976. 

Source: HFEF Survey Data, 1977, PDF p. 180

Figure 6-29: Example of HFEF-N Survey Report from 1977 
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Source: Survey Reports, 1976

Figure 6-30: Example of HFEF-N Air Sample Data Sheet from 1976
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Source: Survey Reports, 1976

Figure 6-31: Example of HFEF-N Smear Sample Data Sheet from 1976
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6.3.2.7 Available Radiological Monitoring Data for ZPPR

Five alpha air monitors were installed in ZPPR to provide for personnel protection. Two of these 

were located in the reactor-air-cooling system (one for each reactor half). These were intended to 

warn personnel of any airborne plutonium particulates from the fuel in the core. The other three alpha 

air monitors were portable and could be moved to optimum locations for personnel protection. One 

monitor was located in-cell to monitor cell air activity. Another was positioned in the workroom. A 

spare portable alpha air monitor was available for use as conditions warranted. Three beta-gamma 

air-particulate monitors were also provided for detection of airborne contamination. One was in the 

reactor cell and the other in the workroom. The third monitor was a spare and could be moved as 

needed (ZPPR FSAR, 1972). 

Table 6-2 shows the trip settings for all particulate air monitors within the ZPPR facility (ZPPR Trip 

Settings, 1969).

Table 6-2: Trip Settings for ZPPR Particulate Air Monitors

Monitor Type Workroom or Cell Upstream Equipment Room Downstream Equipment Room

Alpha N/A 2.0 x 10-8 μCi/cm3 6.0 x 10-12 μCi/cm3

Beta/Gamma 3500 cpm 2.0 x 10-7 μCi/cm3 8.0 x 10-11 μCi/cm3

Seven area-gamma monitors were dispersed in the ZPPR Facility. Two gamma monitors were located 

in the reactor cell, two in the inside equipment room, one in the workroom, one in the core coating 

room, and one at the suspect filters in the fan loft. Three gamma-criticality monitors were provided in 

the mound (reactor cell) area. One was located on top of the loading hood in the workroom, one in the 

storage vault, and one in the core-coating room. All three of these monitors had local and remote 

(ZPPR control room) alarms (ZPPR FSAR, 1972).

Daily surveys were conducted of the ZPPR lunch room, RadCon office, and coffee area. ZPPR cells, 

workroom, control room, counting room, storage areas, office area hallways, and hot and cold change 

areas were surveyed weekly. The vault, AFSR, basement, mock-up building, TLD lab, annex office, 

and annex basement were surveyed monthly (ZPPR Surveys, 1976). Figure 6-32 shows a data sheet 

for a smear sample survey conducted in ZPPR in 1973. 
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Source: ZPPR, 1973

Figure 6-32: Example ZPPR Smear Sample Data Sheet from 1973 

6.3.2.8 Available Radiological Monitoring Data for the L&O Building 

Daily air samples were taken in the area of the Junior Caves, and daily surveys were taken of any 

carpeted areas within the building. Survey frequencies for other areas varied depending on the 

primary function of the area. Chemistry labs, loading docks, and the rear access area and receiving 

areas for the Junior Caves were all surveyed weekly. Twice-monthly surveys were conducted in the 

locker room, cafeteria, Metallurgy Lab, Health Physics office, Junior Cave shop area, and L&O office 

area corridors. All other areas, including interior stairwells, mass spectrometry shop, machine shop, 

and library were surveyed monthly (L&O Surveys, 1969). Figure 6-33 shows a routine smear survey

of the L&O Building Chemistry Labs. Figure 6-34 shows an example survey map for the L&O 

Building Junior Cave Area. 
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Source: L&O Surveys, 1973

Figure 6-33: Example L&O Building Routine Surveys – 1973
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Source: L&O Surveys, 1974

Figure 6-34: Example L&O Building Junior Cave Area Survey Map
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7.0 Feasibility of Dose Reconstruction for the Class Evaluated by 

NIOSH

The feasibility determinations for the class of employees under evaluation in this report are governed 

by both EEOICPA and 42 C.F.R. § 83.13(c)(1). Under that Act and rule, NIOSH must establish 

whether or not it has access to sufficient information either to estimate the maximum radiation dose 

for every type of cancer for which radiation doses are reconstructed that could have been incurred 

under plausible circumstances by any member of the class, or to estimate the radiation doses to 

members of the class more precisely than a maximum dose estimate. If NIOSH has access to 

sufficient information for either case, NIOSH would then determine that it would be feasible to 

conduct dose reconstructions. 

In determining feasibility, NIOSH begins by evaluating whether current or completed NIOSH dose 

reconstructions demonstrate the feasibility of estimating with sufficient accuracy the potential 

radiation exposures of the class. If the conclusion is one of infeasibility, NIOSH systematically 

evaluates the sufficiency of different types of monitoring data, process and source or source term data, 

which together or individually might assure that NIOSH can estimate either the maximum doses that 

members of the class might have incurred, or more precise quantities that reflect the variability of 

exposures experienced by groups or individual members of the class as summarized in Section 7.6. 

This approach is discussed in DCAS’s SEC Petition Evaluation Internal Procedures which are

available at http://www.cdc.gov/niosh/ocas. The next four major subsections of this Evaluation 

Report examine:

• The sufficiency and reliability of the available data. (Section 7.1) 

• The feasibility of reconstructing internal radiation doses. (Section 7.2) 

• The feasibility of reconstructing external radiation doses. (Section 7.3) 

• The bases for petition SEC-00224 as submitted by the petitioner. (Section 7.4) 

7.1 Pedigree of ANL-W Data

This subsection answers questions that need to be asked before performing a feasibility evaluation. 

Data Pedigree addresses the background, history, and origin of the data. It requires looking at site 

methodologies that may have changed over time; primary versus secondary data sources and whether 

they match; and whether data are internally consistent. All these issues form the bedrock of the 

researcher’s confidence and later conclusions about the data’s quality, credibility, reliability, 

representativeness, and sufficiency for determining the feasibility of dose reconstruction. The 

feasibility evaluation presupposes that data pedigree issues have been settled. 
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7.1.1 Internal Monitoring Data Pedigree Review

ATTRIBUTION: Section 7.1.1 was completed by Mike Mahathy and Jason Davis, Oak Ridge 

Associated Universities, Brian Gleckler, Dade Moeller, Inc., and Mitch Findley, MJW Corporation. 

All conclusions drawn from the data regarding the specified facilities were peer-reviewed by the 

individuals listed on the cover page. The rationale for all conclusions in subsequent sections of this 

document devoted to the specified facilities are explained in the associated text. 

Upon request, ANL-W provides database printouts and copies of original hardcopy bioassay results, 

as well as whole-body-count/chest-count records. NIOSH has obtained both bioassay results and a 

limited number of air monitoring results for the period under evaluation. Bioassay sample results 

were obtained starting in 1953. The bioassay results are comprised of urinalyses starting in 1953, and 

fecal results starting in 1961. NIOSH has also obtained whole-body-counting results starting in 1961. 

During the evaluation period, the bioassay records were determined to be incomplete prior to 1958 

due to missing records. Many of the early bioassay records were discovered in the personal files of 

the EBR-I Complex Radiation Protection Manager; although the records were incomplete, they 

included data from important radiological events. 

A large amount of raw bioassay data exists at ANL-W. Many of these records have been compiled 

into a single database, which also contains bioassay data for INL, with a data validation and comment 

field added to each record. These data are in the process of being cross-verified with scanned copies 

of 32,951 pages of internal dosimetry records (Bioassay, 1956-1984; Bioassay, 1956-1967; Bioassay, 

1958-1986). Duplicate records and unusable records (e.g., gross errors) are not included in this 

assessment.

7.1.2 External Monitoring Data Pedigree Review

ATTRIBUTION: Section 7.1.2 was completed by Mike Mahathy and Jason Davis, Oak Ridge 

Associated Universities, Brian Gleckler, Dade Moeller, Inc., and Mitch Findley, MJW Corporation. 

All conclusions drawn from the data regarding the specified facilities were peer-reviewed by the 

individuals listed on the cover page. The rationale for all conclusions in subsequent sections of this 

document devoted to the specified facilities are explained in the associated text. 

Upon request, ANL-W provides database printouts and copies of original hard-copy external 

dosimetry results. NIOSH has obtained personnel dosimetry information and area monitoring data for 

much of the period under evaluation. External dosimetry results were obtained starting in 1952, with 

some breaks in 1955 through 1957. This break appears to be attributable to a change in the site 

responsible for maintaining the external dosimetry results. From a review of external dosimetry 

records of early ANL-W workers, from both ANL-E (Lemont, IL) and INL, the responsibility clearly 

rested with ANL-E through March 1955. After that, there appears to have been a transition of 

responsibility to INL with much more complete records after 1957. A large amount of personnel 

dosimetry data exists at ANL-W, including duplicate records and unusable records (e.g., gross errors). 

NIOSH has determined that all the workers in ANL-W’s radiological areas were monitored for 

external dose starting in 1958 and the dose values exist in original hard-copy form as well as database 

form that are (and have been) provided by DOE for each worker. In addition to the dosimetry data 

that are considered a primary resource for supporting external dose reconstructions, NIOSH also has 

access to ANL-W area monitoring data (including radiation survey data and source-term information) 
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as well as onsite ambient (environmental) data that are also in original hard-copy form. These data 

support the assignment of onsite ambient doses for all monitored ANL workers to account for any 

radiation dose that may have been inappropriately subtracted from their dosimeter results as 

background radiation, as well as the assignment of dose to unmonitored individuals. 

7.1.3 Area Monitoring Data Pedigree Review

ATTRIBUTION: Section 7.1.3 was completed by Mike Mahathy and Jason Davis, Oak Ridge 

Associated Universities, Brian Gleckler, Dade Moeller, Inc., and Mitch Findley, MJW Corporation. 

All conclusions drawn from the data regarding the specified facilities were peer-reviewed by the 

individuals listed on the cover page. The rationale for all conclusions in subsequent sections of this 

document devoted to the specified facilities are explained in the associated text. 

Data from the time period under evaluation consist of area survey reports, contamination smear survey 

reports, and air sampling data. These surveys are copies of original reports and are therefore 

considered primary data sources. However, during the evaluation period it was determined that there 

are known gaps in the area monitoring data prior to 1958. 

7.2 Evaluation of Bounding Internal Radiation Doses at ANL-W 

ATTRIBUTION: Section 7.2 was completed by Mike Mahathy and Jason Davis, Oak Ridge Associated 

Universities, Brian Gleckler, Dade Moeller, Inc., and Mitch Findley, MJW Corporation. All 

conclusions drawn from the data regarding the specified facilities were peer-reviewed by the 

individuals listed on the cover page. The rationale for all conclusions in subsequent sections of this 

document devoted to the specified facilities are explained in the associated text. 

The principal source of internal radiation doses for members of the class under evaluation would have 

been inhalation and ingestion of radiological contamination during the following operations: 

• Destructive and non-destructive nuclear reactor tests

• Irradiation of experimental nuclear reactor fuels

• Irradiation of various materials to transform them by neutron bombardment and/or to study their 

behavior within a nuclear reactor environment 

• Uranium fuel production, which involved work with uranium and thoria 

• Waste management operations supporting all site radiological activities

The potential internal sources would have been dependent on the operational area and activities. The 

major intake sources, as indicated in the bioassay records and incident reports, have been mixed 

fission products (often limited to radioactive iodines and noble gases), activation products, uranium, 

and plutonium. However, the records list a wide spectrum of radionuclides that were monitored and 

an even longer list of codes used to identify the radionuclides, groups of radionuclides, specific 

measurement techniques, or combinations of radionuclides and techniques. Many of the radionuclides 

apply to a small set of workers on a research project, or to workers whose tasks could have exposed 

them to many different sources (e.g., radiation monitoring technicians) (ORAUT-TKBS-0006-5).
SEC-00224 02-18-16 ANL-W 

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Given the large variety of facilities and the ever-changing cast of contractors, the AEC established and 

ran a Health and Safety (H&S) Laboratory at the Idaho site to provide technical support and 

consistency. The organization went through a series of name changes: Health Services Laboratory 

(HSL), Health and Safety Division (HSD), Idaho Center for Radiological and Environmental 

Sciences, and Radiological and Environmental Sciences Laboratory (RESL). During the period under 

evaluation, in-vitro bioassay samples collected throughout the Idaho site were analyzed by the HSL in 

Building CF-646. Beginning in 1963, in-vitro bioassay and WBC facilities were both housed in 

Building CF-690, where a newly-designed counter was used (INEL Bioassay History, 1994). 

Eventually, a mobile whole-body counter was built and periodically brought to the ANL-W facilities, 

reducing the issue of lost work time for WBCs. 

For ANL-W EEOICPA claims, ambient environmental internal doses are only assessed for certain 

unmonitored workers, as discussed in the guidance in ORAUT-TKBS-0007-6, Section 5.6. ORAUTTKBS-0007-4 and its supplement (Peterson, 2004) provide the historical background, rationale, and 

environmental intake data for reconstructing occupational environmental internal doses at the ANL-W 

site. In summary, ORAUT-TKBS-0007-4 provides annual environmental intakes of the significant 

radionuclides for each major operating area on the ANL-W site. Because environmental exposures 

are accounted for in personnel occupational exposures and do not impact the ability to reconstruct 

dose with sufficient accuracy, further assessment of environmental internal exposure is not included in 

this report. 

The following subsections address the ability to bound internal doses, methods for bounding doses, 

and the feasibility of internal dose reconstruction. 

7.2.1 Evaluation of Bounding Internal Radiation Doses for the EBR-I Complex

ATTRIBUTION: Section 7.2.1 was completed by Mike Mahathy, Oak Ridge Associated Universities, 

and Mitch Findley, MJW Corporation. All conclusions drawn from the data regarding the EBR-I 

Complex were peer-reviewed by the individuals listed on the cover page. The rationale for all 

conclusions in subsequent sections of this document devoted to the EBR-I Complex are explained in 

the associated text.

The philosophy of ANL-W’s internal dose monitoring program established a routine monitoring and 

control program based on mixed fission products as the controlling radionuclides. That is, if the 

mixed fission products were controlled to permissible levels, the alpha hazards would be adequately 

controlled. However, potential exceptions to this were noted in Section 5.2.2. 

ANL-W workers participated in bioassay programs for potential intakes of mixed fission products by 

either urinalysis or whole-body counts, either routinely or when workplace indicators (e.g., elevated 

airborne radioactivity, contamination event, positive nasal smear) indicated that an intake may have 

occurred. When workplace indicators indicated that an intake may have occurred, “special” (nonroutine) bioassay would be requested by ANL-W Health Physics. 

As indicated in Section 5.2.1, the categories of radionuclides at the EBR-I Complex included: mixed 

fission and activation products, radioiodines, plutonium, uranium, thorium, and other radionuclides. 
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7.2.1.1 Evaluation of Bounding Process-Related Internal Doses for the EBR-I Complex

NIOSH obtained routine bioassay data for EBR-I Complex workers for some of the period under 

evaluation. NIOSH obtained routine urinalysis data starting in 1958 and in-vivo data starting in 1962. 

NIOSH obtained some bioassay results (both urinalysis and fecal) for the period 1952 through 1957, 

but those bioassay results appear to be incomplete and possibly such monitoring was not based on 

routine sampling. Incident-driven in-vitro bioassay following radiological events of concern were 

discovered during the evaluation period. For example, there was extensive bioassay for workers after 

the discovery of a ruptured blanket brick in April 1955, and after the partial core meltdown in 

November 1955. 

The principal sources of internal radiation doses for members of the class working at the EBR-I, 

ZPR-III, AFSR, and the BORAX-I, -II, -III, -IV, and -V reactors include uranium and mixed fission 

products; thorium was a potential source of exposure with BORAX-IV as well. The following 

subsections address the ability to bound internal doses, methods for bounding doses, and the 

feasibility of internal dose reconstruction. 

BORAX-I – BORAX-V and AFSR

Bioassay Information and Available Data 

As previously discussed in Section 6.1, there is significant programmatic information about the 

ANL-W’s bioassay program, and after 1957, there are extensive in-vitro and in-vivo bioassay data that 

NIOSH believes are sufficient for bounding internal radiation doses from EBR-I Complex 

radionuclides. The majority of the available bioassay data include: gross beta radioactivity in urine, 

gross gamma radioactivity in urine, Sr-90 in urine, and whole-body count results. 

Not all of the 50 workers referenced in Table 6-1 were employed at ANL-W for the entire period from 

1952 through 1960. Table 7-1 presents the internal monitoring data available, by analysis and 

radiation type, for 12 of the 50 workers known to have worked at ANL-W for the entire nine-year 

period. 

NOTE: Data for nine of the twelve workers are available for 1952; however, those results are not 

included as it is not clear whether or not they represented baseline, routine, or incident 

sampling; thus, the analysis below covers the eight years from 1953 through 1960. 

Table 7-1 was compiled using data obtained directly from INL and ANL in response to record 

requests, and data identified during data captures (including bioassay lab results cards). In 1953, none 

of the workers was monitored. In 1954, five workers were monitored for uranium intakes. In 1955, 

all 12 were monitored for intakes of uranium and/or plutonium. NIOSH has found bioassay data for 

only two of the 12 workers for 1956 and no monitoring data for 1957. In 1958, 11 workers were 

monitored for intakes of beta-emitting radiation and two were monitored for uranium intakes. For 

1959, NIOSH has found bioassay data for only two of the 12 workers. For 1960, NIOSH has found 

bioassay data for 11 of the workers; gross beta results are available for all 11 while gross gamma is 

available for one. The occupation of worker No. 7 was [job category redacted]. 

NIOSH believes bioassay data available for the period 1953 through 1957 are inadequate to bound 

radionuclide intakes. NIOSH has found a small number of bioassay results for 1959. NIOSH intends 
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to use bioassay data available for 1958 and 1960 to interpolate intakes and doses for 1959. All 

bioassay data obtained for 1955 resulted from incident follow-ups, not routine monitoring. 

Table 7-1: Internal Monitoring Data from All Sources for 12 Selected Workers (1953-1960)

Worker 1953 1954 1955 1956 1957 1958 1959 1960

1 No data 

provided

urine 

U-235, 

fecal 

U-235

urine 

U-235, 

urine Pu

urine beta No data 

provided urine beta No data 

provided urine beta

2 No data 

provided

No data 

provided

urine 

U-235, 

urine Pu

urine beta No data 

provided

urine beta, 

urine 

U-235

No data 

provided urine beta

3 No data 

provided

urine 

U-235

urine 

U-235, 

urine Pu

No data 

provided

No data

provided urine beta urine beta urine beta, 

gamma

4 No data 

provided

No data 

provided

urine 

U-235

No data 

provided

No data 

provided urine beta No data provided urine beta

5 No data 

provided

fecal 

U-235

urine 

U-235, 

urine Pu

No data 

provided

No data 

provided

urine beta, 

urine 

U-235

No data 

provided urine beta

6 No data 

provided

urine 

U-235

urine 

U-235

No data 

provided

No data 

provided urine beta No data provided urine beta

7 No data 

provided

No data 

provided

urine 

U-235

No data 

provided

No data 

provided

No data 

provided

No data 

provided

No data 

provided

8 No data 

provided

No data 

provided

urine 

U-235

No data 

provided

No data 

provided urine beta No data provided urine beta

9 No data 

provided

urine beta, 

urine 

U-235

urine 

U-235, 

urine Pu

No data 

provided

No data 

provided

urine beta, 

urine 

U-235

No data 

provided urine beta

10 No data 

provided

No data 

provided

urine 

U-235

No data 

provided

No data 

provided urine beta No data provided urine beta

11 No data 

provided

urine beta, 

urine 

U-235

urine 

U-235, 

urine Pu

No data 

provided

No data 

provided urine beta urine beta urine beta

12 No data 

provided

No data 

provided

urine 

U-235

No data 

provided

No data 

provided urine beta No data provided urine beta

Green indicates alpha radiation monitoring. 

Blue indicates beta/gamma radiation monitoring. 

Only data received directly from INL and ANL in response to records requests were used to populate 

Table 7-2. Had additional data not been located during data capture, NIOSH would not know workers 

were monitored for intakes in 1955 and some in 1958, 1959, and 1960. 
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Table 7-2: Internal Monitoring Data Provided by Sites for 12 Selected Workers (1953-1960)

Worker 1953 1954 1955 1956 1957 1958 1959 1960

1 No data 

provided

urine 

U-235, 

fecal 

U-235

No data 

provided urine beta No data 

provided urine beta No data 

provided urine beta

2 No data 

provided

No data 

provided

No data 

provided urine beta No data 

provided

No data 

provided

No data 

provided

No data 

provided

3 No data 

provided

urine 

U-235

No data 

provided

No data

provided

No data 

provided urine beta urine beta

urine 

beta, 

gamma

4 No data 

provided

No data 

provided

No data 

provided

No data 

provided

No data 

provided urine beta No data provided urine beta

5 No data 

provided

fecal 

U-235

No data 

provided

No data 

provided

No data 

provided

urine 

beta, 

urine 

U-235

No data 

provided urine beta

6 No data 

provided

urine 

U-235

No data 

provided

No data 

provided

No data 

provided urine beta No data provided

No data 

provided

7 No data 

provided

No data 

provided

No data 

provided

No data 

provided

No data 

provided

No data 

provided

No data 

provided

No data 

provided

8 No data 

provided

No data 

provided

No data 

provided

No data 

provided

No data 

provided urine beta No data provided urine beta

9 No data 

provided

urine beta, 

urine 

U-235

No data 

provided

No data 

provided

No data 

provided

No data 

provided No data 

provided

No data 

provided

10 No data 

provided

No data 

provided

No data 

provided

No data 

provided

No data 

provided

No data 

provided

No data 

provided

No data 

provided

11 No data 

provided

urine beta, 

urine 

U-235

No data 

provided

No data 

provided

No data 

provided

No data 

provided No data 

provided

No data 

provided

12 No data 

provided

No data 

provided

No data 

provided

No data 

provided

No data 

provided

No data 

provided

No data 

provided

No data 

provided

Green indicates alpha radiation monitoring. 

Blue indicates beta/gamma radiation monitoring. 

Using external dosimetry reports, NIOSH assembled a roster of workers at AFSR from 1961 through 

1970. NIOSH researched bioassay results available for each worker. Table 7-3 presents the number 

of AFSR workers monitored by method each year and the total number of workers with no bioassay 

results. Only two workers were not monitored in 1961 and one each in 1968 and 1969. In each 

instance, the occupation of said workers was [job category redacted]; each of the four workers was 

monitored in the subsequent year. Table 7-4 presents similar information for BORAX-V workers 

through the operation of that reactor from 1961 through 1964. Results are available for each of the 

BORAX-V workers, indicating they were all routinely monitored. 
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Table 7-3: Monitoring of AFSR Workers

Worker Status 1961 1962 1963 1964 1965 1966 1967 1968 1969 1970

Total Workers 9 10 9 8 8 9 9 9 7 6

Workers with In-Vitro Results 7 10 8 8 8 6 2 2 0 0 

Workers with In-Vivo Results 0 0 6 2 1 6 9 7 6 6

Workers with No Results 2 0 0 0 0 0 0 1 1 0 

Table 7-4: Monitoring of BORAX-V Workers

Worker Status 1961 1962 1963 1964

Total Workers 28 28 34 28

Workers with In-Vitro Results 28 28 34 28

Workers with In-Vivo Results 0 2 17 5 

Workers Not Monitored 0 0 0 0 

Routine bioassay for BORAX-V and AFSR workers is available starting in 1961, as shown in Tables 

7-3 and 7-4. The earliest in-vivo data for BORAX-V workers in the available data are for 1962. 

Starting with BORAX-V in 1962, NIOSH has obtained urinalysis and/or in-vivo results for all workers 

on external monitoring rosters for the reactor. 

From 1958 through 1961, urinalysis results will be used to bound doses received from alpha, beta and 

X-ray, and gamma-emitting radiation. While such doses would be very small, NIOSH will use a ratio 

of uranium to fission products to derive doses received from uranium in 1958 when BORAX-IV 

conducted cladding failure experiments (Borax-IV Test, 1959). Starting with 1963, the in-vivo data 

will be used with urinalysis results to bound doses received from X-ray and gamma-emitting 

radiation. 

Air Monitoring Information and Available Data

As previously discussed in Section 6.3, there is significant programmatic information about the 

ANL-W’s air monitoring programs, and there are extensive air sampling data that NIOSH believes are 

sufficient for bounding internal radiation doses from the radionuclides if they could not be bounded 

using the available bioassay data (e.g., thorium, uranium without fission products). However, the air 

monitoring data prior to 1958 are incomplete. While records exist indicating that air monitoring was 

performed and the data collected and analyzed, the data are limited at best. The primary source of 

pre-1958 air monitoring data was from the personal files of the EBR-I Complex Radiological 

Protection Manager. ANL-W did not routinely monitor for airborne radioactivity at AFSR due to the 

solid structure of the small lab-scale reactor. The majority of the available air sampling data includes 

both gross beta-gamma radioactivity and gross alpha radioactivity results for each air sample.
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Alternative Data Sources for Bounding Internal Dose 

NIOSH has found little bioassay data for 1959 for the EBR-I complex but results for 1959 will be 

interpolated using results obtained for 1958 and 1960. 

As previously discussed in Section 6.3, there is significant programmatic information about the 

ANL-W’s contamination monitoring programs, and there are extensive contamination smear data that 

could be used to support bounding potential doses to certain radionuclides. Although available, 

NIOSH does not have any current plans to use these data. 

7.2.1.2 Methods for Bounding Operational Period Internal Doses for the EBR-I Complex

Fission and Activation Products 

Beginning in 1958, potential intakes of mixed fission and activation products for monitored and 

unmonitored EBR-I Complex workers can be estimated using the approaches described in ORAUTTKBS-0007-5. For urine samples analyzed for gross beta, gross gamma, and/or strontium 

radioactivity, NIOSH will assess Sr-90 and/or Cs-137 intakes in accordance with ORAUT-TKBS0007-5 and ORAUT-OTIB-0060. For whole body counts, NIOSH will assess Cs-137 intakes in 

accordance with ORAUT-TKBS-0007-5 and ORAUT-OTIB-0060. In accordance with the guidance 

in ORAUT-OTIB-0054, NIOSH will use Sr-90 and/or Cs-137 intakes to estimate the intakes of the 

other mixed fission and activation products that may have been present, unless a bioassay result was 

available for a specific radionuclide. In the event that a bioassay result was available for a specific 

radionuclide, NIOSH will use that result in lieu of the ORAUT-OTIB-0054 approach. 

Based on the procedural information and the data on-hand, NIOSH finds that it has adequate 

monitoring data to allow for sufficiently accurate estimation of internal doses attributable to mixed 

fission and activation products for EBR-I Complex workers at the EBR-I Complex from 1958 

forward. Such doses cannot be bounded from 1952 through 1957 due to the lack of adequate 

monitoring data, as demonstrated in Table 7-1. 

Radioiodines 

Based on the availability of gross-gamma urine-sample data and whole-body count data for EBR-I 

Complex workers, NIOSH finds that it has adequate monitoring data to allow for sufficiently accurate 

estimation of internal doses attributable to radioiodines from 1958 forward. Such doses cannot be 

bounded from 1952 through 1957 due to the lack of adequate monitoring data, as demonstrated in 

Table 7-1. 

The potential intakes of radioiodines without other mixed fission products present can be estimated 

for monitored EBR-I Complex workers using the bioassay data and guidance provided in Section 

5.5.4 of ORAUT-TKBS-0007-5. 

When radioiodines are present with other mixed fission products, the potential intakes of radioiodines 

for monitored EBR-I Complex workers can be accounted for using the bioassay data and guidance 

provided in Section 5.5.1 of ORAUT-TKBS-0007-5, which employs the use of ORAUT-OTIB-0054. 
SEC-00224 02-18-16 ANL-W 

197 of 260 

Based on the available information, the presence of airborne radioiodine without other mixed fission 

products present was uncommon at the EBR-I Complex, and unmonitored intakes of radioiodines 

likely only occurred when other mixed fission products were present. Therefore, potential intakes of 

radioiodines for unmonitored EBR-I Complex workers can be estimated using the guidance provided 

in Section 5.6 of ORAUT-TKBS-0007-5. 

Radioactive Noble Gases

ANL-W was not involved in the mining, milling, or separation of uranium or thorium from ores. 

Fuels arrived at the site pre-formed and extruded and did not age to the point where radon or thoron 

would have been a significant concern. Therefore, radon and thoron are not considered a significant 

internal exposure concern. Because fission-produced radioactive noble gases do not represent a 

significant internal exposure concern, NIOSH finds there is no need to assess the internal doses 

attributable to radioactive noble gases.

Actinides with Mixed Fission Products Present 

With a few exceptions, the majority of the actinides at the EBR-I Complex were present with mixed 

fission products. The potential intakes of actinides when mixed fission products were present can be 

estimated using the approach described in Section 5.5.2 of ORAUT-TKBS-0007-5. Based on that 

approach, actinide intakes are estimated by multiplying the assessed Sr-90 and/or Cs-137 intakes by 

actinide-to-Sr-90 and/or actinide-to-Cs-137 ratios. 

Based on the procedural information and the data on hand, NIOSH finds that it has adequate 

monitoring data to allow for sufficiently accurate estimation of internal doses attributable to actinides 

when mixed fission products were present for EBR-I Complex workers during the entire period that 

radiological operations took place at the EBR-I Complex when sufficient monitoring data are 

available.

Thorium was a concern only at the BORAX-IV reactor although thoria was sealed in fuel elements in 

the reactor core. There was the potential for very small intakes of thorium resulting from release of 

radioactivity with the BORAX-IV fuel failure testing conducted in 1958; however, no alpha radiation 

was detected from monitoring conducted during the events and post-events experiments. Potential 

thorium intakes during the testing conducted in 1958 can be bounded using thorium-to-fission-product 

ratios (Borax-IV Test, 1959). 

Uranium without Mixed Fission Products Present 

The potential for uranium exposures was minimal because encapsulated fuel sources were used at the 

EBR-I Complex reactors. EBR-I Health Physics used air monitoring data and/or contamination smear 

data to control potential uranium exposure at these reactors. There was potential for very small 

intakes of uranium resulting from release of radioactivity during the BORAX-IV fuel-failure testing 

conducted in 1958; however, no alpha radiation was detected from monitoring conducted during the 

events and post-events experiments. Uranium intakes during the 1958 testing can be bounded using 

uranium equivalence of fission-gas release ratios derived from fuel analysis given in reference: 

Borax-IV Test, 1959. A small potential existed for exposure to depleted uranium used as blanket 

material in the ZPR-III mock-ups. However, the low-specific-activity of the material, small 

contamination levels when discovered, and rigorous contamination monitoring practices at ZPR-III 
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make those exposures highly unlikely. Because there are inadequate uranium bioassay data to 

estimate the potential intakes of uranium without mixed fission products at the EBR-I Complex after 

1957, NIOSH will bound the potential intakes of uranium by assigning uranium intakes based on 10% 

of the maximum permissible concentration (MPC) that ANL-W was using. A review of the available 

air monitoring data indicated that the airborne alpha radioactivity at the EBR-I Complex was typically 

below 10% of the MPC. Because it cannot be determined when and where the workers within the 

EBR-I Complex worked, these uranium intakes will be assessed for all workers any time they were 

within the EBR-I Complex after 1957. 

Plutonium without Mixed Fission Products Present 

ZPR-III plutonium fuel plates and the Mark III core of EBR-I represent the major sources of 

plutonium during operations at the EBR-I complex. With the exception of contamination surveys of 

the initial plutonium fuel plates for ZPR-III, there is no evidence of plutonium contamination while 

plutonium fuel plates were used. Likewise, the third core at EBR-I was a plutonium core but 

functioned without a breach in fuel rod cladding throughout its use. At the EBR-I Complex, there are 

no known exposure scenarios to plutonium without mixed fission products being present. 

7.2.2 Evaluation of Bounding Internal Radiation Doses for the EBR-II Complex

ATTRIBUTION: Section 7.2.2 was completed by Jason Davis, Oak Ridge Associated Universities, and 

Brian Gleckler, Dade Moeller, Inc. All conclusions drawn from the data regarding the EBR-II 

Complex were peer-reviewed by the individuals listed on the cover page. The rationale for all 

conclusions in subsequent sections of this document devoted to the EBR-II Complex are explained in 

the associated text.

The philosophy of ANL-W’s internal dose monitoring program established a routine monitoring and 

control program based on mixed fission products as the controlling radionuclides. That is, if the 

mixed fission products were controlled to permissible levels, the alpha hazards would be adequately 

controlled. However, potential exceptions to this were noted in Section 5.2.2 

ANL-W workers participated in bioassay programs for potential intakes of mixed fission products by 

either urinalysis or whole-body counts, either routinely or when workplace indicators (e.g., elevated 

airborne radioactivity, contamination event, positive nasal smear) indicated that an intake may have 

occurred. When workplace indicators indicated that an intake may have occurred, “special” (nonroutine) bioassay would be requested by ANL-W Health Physics. 

7.2.2.1 Evaluation of Bounding Process-Related Internal Doses for the EBR-II Complex

The principal sources of internal radiation doses for members of the evaluated class working at the 

TREAT, L&O Building, and ZPPR facilities include mixed fission and activation products, 

radioiodines, radioactive noble gases, plutonium, uranium, thorium, other actinides, and other 

radionuclides. The following subsections address the ability to bound internal doses, methods for 

bounding doses, and the feasibility of internal dose reconstruction. 
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Bioassay Information and Available Data 

As previously discussed in Section 6.1, there is significant programmatic information about the 

ANL-W’s bioassay program, and after 1957, there are extensive in-vitro and in-vivo bioassay data that 

NIOSH believes are sufficient for bounding internal radiation doses from most EBR-II Complex 

radionuclides. The majority of the available bioassay data include: gross beta radioactivity in urine, 

gross gamma radioactivity in urine, Sr-90 in urine, and whole-body count results. 

Air Monitoring Information and Available Data 

As previously discussed in Section 6.3, there is significant programmatic information about the 

ANL-W’s air monitoring programs, and there are extensive air sampling data that NIOSH believes are 

sufficient for bounding internal radiation doses from the radionuclides that could not be bounded 

using the available bioassay data (e.g., thorium, uranium without fission products, etc.). The majority 

of the available air sampling data includes both gross beta-gamma radioactivity and gross alpha 

radioactivity results for each air sample. In addition, a significant percentage of the air samples were 

counted more than once for gross alpha radioactivity, and the subsequent results were often lower, 

suggesting that the bulk of the detected activity was due to short-lived, naturally-occurring radon 

decay products. 

Alternative Data Sources for Bounding Internal Dose 

As previously discussed in Section 6.3, there is significant programmatic information about the 

ANL-W’s contamination monitoring programs, and there are extensive contamination smear data that 

could be used to support bounding potential doses to certain radionuclides. Although available, 

NIOSH does not have any current plans to use these data. 

7.2.2.2 Methods for Bounding Operational Period Internal Doses for the EBR-II Complex

Fission and Activation Products 

The potential intakes of mixed fission and activation products for monitored and unmonitored EBR-II 

Complex workers can be estimated using the approaches described in ORAUT-TKBS-0007-5. For 

urine samples analyzed for gross beta, gross gamma, and/or strontium radioactivity, NIOSH will 

assess Sr-90 and/or Cs-137 intakes in accordance with ORAUT-TKBS-0007-5 and ORAUT-OTIB0060. For whole-body counts, NIOSH will assess Cs-137 intakes in accordance with ORAUT-TKBS0007-5 and ORAUT-OTIB-0060. In accordance with the guidance in ORAUT-OTIB-0054, NIOSH 

will use Sr-90 and/or Cs-137 intakes to estimate the intakes of the other mixed fission and activation 

products that may have been present, unless a bioassay result is available for a specific radionuclide. 

In the event that a bioassay result is available for a specific radionuclide, NIOSH will use that result in 

lieu of the ORAUT-OTIB-0054 approach. 

Based on the procedural information and the data on hand, NIOSH finds that it has adequate 

monitoring data to allow for sufficiently accurate estimation of internal doses attributable to mixed 

fission and activation products for EBR-II Complex workers during the entire period that radiological 

operations took place at the EBR-II Complex.
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Radioiodines 

Based on the availability of the gross gamma urine sample data and whole-body count data for EBR-II 

Complex workers, NIOSH finds that it has adequate monitoring data to allow for sufficiently accurate 

estimation of internal doses attributable to radioiodines for all periods at the EBR-II Complex. 

The potential intakes of radioiodines without the other mixed fission products present can be 

estimated for monitored EBR-II Complex workers using the bioassay data and guidance provided in 

Section 5.5.4 of ORAUT-TKBS-0007-5. When the radioiodines are present with the other mixed 

fission products, the potential intakes of radioiodines for monitored EBR-II Complex workers can be 

accounted for using the bioassay data and guidance provided in Section 5.5.1 of ORAUT-TKBS0007-5, which employs the use of ORAUT-OTIB-0054. 

Based on the available information, the presence of airborne radioiodine without the other mixed 

fission products present was uncommon at the EBR-II Complex, and unmonitored intakes of 

radioiodines likely only occurred when the other mixed fission products were present. Therefore, 

potential intakes of radioiodines are estimated for unmonitored EBR-II Complex workers using the 

guidance provided in Section 5.6 of ORAUT-TKBS-0007-5. 

Radioactive Noble Gases

ANL-W was not involved in the mining, milling, or separation of uranium or thorium from ores. 

Fuels arrived at the side pre-formed and extruded and did not age to the point where radon or thoron 

would have been a significant concern. Therefore, radon and thoron are not considered a significant 

internal exposure concern. Because radioactive noble gases do not represent a significant internal 

exposure concern, NIOSH finds there is no need to assess the internal doses attributable to radioactive 

noble gases. 

Actinides with Mixed Fission Products Present 

With a few exceptions, the majority of the actinides at the EBR-II Complex were present with mixed 

fission products. The potential intakes of actinides when mixed fission products were present can be 

estimated using the approach described in Section 5.5.2 of ORAUT-TKBS-0007-5. Based on that 

approach, actinide intakes are estimated by multiplying the assessed Sr-90 and/or Cs-137 intakes by 

actinide-to-Sr-90 and/or actinide-to-Cs-137 ratios. 

Based on the procedural information and the data on hand, NIOSH finds that it has adequate 

monitoring data to allow for sufficiently accurate estimation of internal doses attributable to actinides 

when mixed fission products were present for EBR-II Complex workers during the entire period that 

radiological operations took place at the EBR-II Complex. 

Uranium without Mixed Fission Products Present 

The potential for uranium exposures within ZPPR was minimal because encapsulated and coated fuel 

sources were used (SRDB 138139). ZPPR Health Physics used air monitoring data and/or 

contamination smear data to control potential uranium exposure at this reactor during normal 

operations. Loading and unloading procedures were carefully monitored to detect any loose 
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contamination on fuel plates. Exposures to uranium without mixed fission products can be bound 

using 10% of MPC (maximum permissible concentration) from available air monitoring data. 

Because of FCF Hot-Line start-up activities, FCF machine shop activities, and Cold-Line fuel 

production at the FCF, ITF, and FASB, more-than-incidental intakes of depleted and enriched 

uranium without mixed fission products present could have occurred from August 1967 to as late as 

1994. For the period of August 1967 through June 1983, no uranium bioassay data could be found for 

ANL-W workers. Beginning in July 1983, there is a significant increase in uranium bioassays for 

ANL-W workers, and the bioassay data can be used to estimate the intakes and doses from these 

uranium exposures. 

Because there are inadequate uranium bioassay data to estimate the potential intakes of depleted and 

enriched uranium without mixed fission products during the period of August 1967–June 1983, 

NIOSH will estimate those intakes using the gross alpha radioactivity air sampling results for those 

areas. In the instances where the air samples were counted for alpha radioactivity more than once due 

to the presence of short-lived alpha-emitting radionuclides, the latest result for gross alpha 

radioactivity will be used for this approach, since U-234, U-235, and U-238 are long-lived 

alpha-emitting radionuclides. Because it cannot be determined when and where the workers within 

the EBR-II Complex worked, these uranium intakes will be assessed for all workers any time they 

were within the EBR-II Complex during the period of August 1967–June 1983. For periods that 

workers were only assigned EBR-I Complex dosimeters, no additional uranium intakes will be 

assessed using this approach.

As a simplification and because the potential exposures to depleted uranium and enriched uranium 

were intermixed, all alpha radioactivity in the air sample results, along with the resulting inhalation 

intake estimates, will be assessed as 100% U-234. For the inhalation rate, a rate of 1.2 m3

/hr (9.6 m3

per 8 hr workday) will be assumed (ICRP 66, 1994). Because air sampling data can only be used to 

estimate potential intakes due to inhalation, potential intakes due to ingestion will be estimated based 

on the recommendations in the Technical Information Bulletin: Estimation of Ingestion Intakes 

(OCAS-TIB-009). Because the uranium that was processed at the EBR-II Complex may have been 

recycled uranium, and because the Y-12 Plant was the likely source of the high-enriched uranium 

metal feedstock being used to produce the EBR-II Reactor fuel, intakes of the potential radioactive 

impurities in the uranium will be calculated by multiplying the inhalation and ingestion uranium 

intakes by the values in Table 7-5. 

Table 7-5: Default Ratios for Recycled Uranium Intakes

Radionuclide Default ratio 

(pCi/pCi total U)

Tc-99 1.3E-02

Th-228 5E-04

Np-237 6E-04

Pu-238 2E-04

Note: The values in this table are based on the values provided in Table 5-8 of the Y-12 Plant 

Occupational Internal Dose TBD (ORAUT-TKBS-0014-5). The values have also been adjusted in 

accordance with that TBD for use on all types of cases, including best estimate cases.
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Figure 7-1 provides an example of a 1969 air sample result for the FCF Cold-Line. During the 

collection period for this sample, some uranium was spilled from the melt-refining process and 

cleaned up in Room 26 (Survey Reports, July1969).

NOTE: The potential uranium exposures and available Cold-Line air monitoring for the period 

of January 1970 – January 1972 were lower because no Cold-Line fuel pin and fuel 

element production took place at ANL-W during that period. At the end of December 

1969, the Cold-Line fuel pin and fuel element production equipment in FCF Rooms 20 

and 26 was put on stand-by (Progress Report, Dec1969, PDF p. 88). However, the 

Cold-Line continued to fabricate unirradiated subassemblies from vendor-produced fuel 

elements (at ITF) and to perform impact-bonding of the unbonded vendor fuel elements 

(in FCF Room 23) (Progress Report, Dec1969, PDF p. 88). The FCF’s Cold-Line fuel 

pin and fuel element production was likely shut down due to a large backlog of 

completed fuel elements, and given that ANL-W was beginning to receive large 

numbers of acceptable vendor- produced fuel elements. In January 1970, there were a 

total of 23,766 completed fuel elements ready to be put into EBR-II Reactor 

subassemblies (Progress Report, Jan1970, PDF p. 78). In the mid-1970s, the FCF’s 

Cold-Line fuel pin and fuel element production equipment was deactivated (Batte, 1986, 

PDF p. 9). Fuel pin fabrication in FCF Room 20 was not re-initiated until 1982 (Batte, 

1986, PDF p. 9). Given that Cold-Line fuel production in FASB did not begin until 

February 1972, no Cold-Line fuel pin and fuel element production took place at ANL-W 

during the period of January 1970 – January 1972. 
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Source: Survey Reports, July1969

Figure 7-1: Example FCF Cold-Line Air Sample Result from 1969
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Thorium without Mixed Fission Products Present 

During the period of FCF Hot-Line fuel production (August 1963–November 1967), more than 

incidental intakes of thorium without mixed fission products present could have occurred in FCF 

Room 25 due to the use of thoria to coat crucibles and molds for the FCF’s Hot-Line. By December 

1967, the use of thoria as a refractory coating was discontinued, which eliminated the potential for 

thoria intakes after November 1967. Because there are inadequate thorium bioassay data to estimate 

the potential intakes of thorium without mixed fission products during the period of August 1963–

November 1967, NIOSH will bound the potential intakes of thorium by assigning thorium intakes 

based on 10% of the maximum permissible concentration (MPC) that ANL-W was using. A review 

of the available air monitoring data indicated that the airborne alpha radioactivity in FCF Room 25 

was typically below 10% of the MPC. Because it cannot be determined when and where the workers 

within the EBR-II Complex worked, these thorium intakes will be assessed for all workers any time 

they were within the EBR-II Complex during the period of August 1963–November 1967. 

Because nearly all thorium deposits contain some uranium, Th-230 from the uranium series will be 

extracted with the other thorium isotopes. Therefore, any thorium will consist of the isotopes of 

Th-232, Th-228, and Th-230. Between Th-232, Th-228, and Th-230, Th-232 has the highest internal 

dose coefficient for all organs. Therefore, 100% of the gross alpha radioactivity in the air sample 

results will be attributed to Th-232 for the internal dose calculations. Inhalation intakes will be 

calculated using an inhalation rate of 1.2 m3

/hr (9.6 m3 per 8 hr workday) (ICRP 66, 1994). Because 

air sampling data can only be used to estimate potential intakes due to inhalation, potential intakes due 

to ingestion will be estimated based on the recommendations in the Technical Information Bulletin: 

Estimation of Ingestion Intakes (OCAS-TIB-009). 

Figure 7-2 provides an example of a 1963 air sample result for FCF Room 25. During the collection 

period for this sample, some thoria was spilled on the floor from one of the hoods in Room 25 (FCF 

Survey Reports, 1963).
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Source: FCF Survey Reports, 1963

Figure 7-2: Example FCF Room 25 Air Sample Result from 1963

Plutonium without Mixed Fission Products Present 

The potential for plutonium exposures within ZPPR was minimal because encapsulated and coated 

fuel sources were used (SRDB 138139). ZPPR Health Physics used air monitoring data and/or 

contamination smear data to control potential plutonium exposure at this reactor during normal 

operations. Loading and unloading procedures were carefully monitored to detect any loose 
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contamination on fuel plates. Exposures to plutonium without mixed fission products can be bound 

using 10% of MPC (maximum permissible concentration) from available air monitoring data.

Intermittent work associated with the Mark-II type RAS-TREAT sodium-loop experiments at the FCF 

could have resulted in potential intermittent exposures to plutonium without mixed fission products 

present when that work included the removal of the test sections and when that work was performed 

outside of the Air and Argon Cells. As indicated in Section 5.1.2.3, the test section removal work 

took place outside of the FCF hot cells during the period of April 1970–December 1972, and was 

performed mostly in FCF Rooms 22 and 27. After 1972, RAS-TREAT sodium-loop experiment work 

appears to have been performed exclusively in the Argon Cell, and the experiments were irradiated to 

much higher levels. Therefore, potential plutonium exposures after 1972 would have occurred with 

mixed fission products being present (Cook, 1975).

Because there isn’t adequate plutonium bioassay data to estimate the potential intakes of plutonium 

without mixed fission products during the period of April 1970–December 1972, NIOSH will estimate 

those intakes using the gross alpha radioactivity air sampling results that were collected during 

RAS-TREAT sodium-loop experiment work in FCF Rooms 22 and 27. In the instances where the air 

samples were counted for alpha radioactivity more than once due to the presence of short-lived 

alpha-emitting radionuclides, the latest result for gross alpha radioactivity will be used for this 

approach, since isotopes of plutonium are long-lived alpha-emitting radionuclides. Because it cannot 

be determined when and where the workers within the EBR-II Complex worked, these plutonium 

intakes will be assessed for all workers any time they were within the EBR-II Complex during the 

period of April 1970–December 1972. 

The plutonium-bearing Mark-II type loops that were handled at the FCF contained experimental 

mixed oxide (MOX) fuels that were mostly enriched UO2 with a lesser quantity of PuO2 (FCF Rad 

Survey, 1970, PDF pp. 208-211; Cook, 1975; Dickerman, 1979). The plutonium in these fuels was 

predominately Pu-239. Because Pu-239 has a significantly shorter half-life than the isotopes in 

enriched uranium, the alpha radioactivity in these experimental uranium-plutonium fuels would have 

been dominated by plutonium, even though more uranium was present in terms of mass. Given that, 

and because the exact composition of each of the experimental fuels is uncertain, 100% of the gross 

alpha activity in the air samples will be assumed to be Pu-239. The assumption of 100% of the alpha 

radioactivity in the air samples being attributed to Pu-239 would result in a bounding dose for 

exposures to these experimental fuels. 

Using the records already available to NIOSH, if it can be determined when each plutonium-bearing 

loop experiment test section was handled in FCF Rooms 22 and 27, acute plutonium intakes for each 

specific period of that work will be estimated from the available air sample data. Otherwise, chronic 

intakes will be assessed as a bounding approach. Inhalation intakes will be calculated using an 

inhalation rate of 1.2 m3

/hr (9.6 m3 per 8 hr workday) (ICRP 66, 1994). Because air sampling data 

can only be used to estimate potential intakes due to inhalation, potential intakes due to ingestion will 

be estimated based on the recommendations in the Technical Information Bulletin: Estimation of 

Ingestion Intakes (OCAS-TIB-009). 
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Figure 7-3 provides an example of a 1970 air sample result for FCF Room 27 that was collected 

during some RAS-TREAT sodium-loop experiment work (Survey Reports, Aug1970).

Source: Survey Reports, Aug1970, PDF p. 111

Figure 7-3: Example Air Sample Result for FCF RAS-TREAT Sodium-Loop Work from 1970
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Other Radionuclides 

Even though EBR-II Complex workers rarely received bioassay measurements for H-3 and/or C-14 

during the period of 1949–1979, these two radionuclides were likely only present where mixed fission 

and activation products were present. Furthermore, the source terms for H-3 and C-14 and their 

ability to contribute to the workers’ internal doses were likely insignificant compared to the workers’ 

exposures to mixed fission and activation products. 

The potential intakes of Na-24 can be estimated for monitored and unmonitored EBR-II Complex 

workers using the guidance provided in Section 5.5.8 of ORAUT-TKBS-0007-5. Urine samples 

analyzed for gross gamma radioactivity and whole-body counts can be used to assess any potential 

intakes of Na-24. Based on the availability of the gross beta urine sample data, gross gamma urine 

sample data, and whole-body count data for EBR-II Complex workers, NIOSH finds that it has 

adequate monitoring data to allow for sufficiently accurate estimation of internal doses attributable to 

Na-24 for EBR-II Complex workers the entire period that radiological operations took place at the 

EBR-II Complex. 

7.2.3 Internal Dose Reconstruction Feasibility Conclusion 

ATTRIBUTION: Section 7.2.3 was completed by Mike Mahathy and Jason Davis, Oak Ridge 

Associated Universities, Brian Gleckler, Dade Moeller, Inc., and Mitch Findley, MJW Corporation. 

All conclusions drawn from the data regarding the specified facilities were peer-reviewed by the 

individuals listed on the cover page. The rationale for all conclusions in subsequent sections of this 

document devoted to the specified facilities are explained in the associated text. 

NIOSH has come to the following conclusions regarding the feasibility of internal dose reconstruction 

for the EBR-I Complex facilities and EBR-II Complex facilities under evaluation for this report: 

• NIOSH concludes that, due to missing records, the available bioassay data are such that internal 

radiation doses received from intakes of fission products and actinides cannot be completely 

reconstructed with sufficient accuracy from April 10, 1951 through December 31, 1957. 

• NIOSH concludes that the available bioassay data are such that internal radiation doses received 

from intakes of fission products and actinides can be completely reconstructed with sufficient 

accuracy from January 1, 1958 through December 31, 1979. 

Although NIOSH found that it is not possible to completely reconstruct internal radiation doses for the 

period from April 10, 1951 through December 31, 1957, NIOSH intends to use any internal 

monitoring data that may become available for an individual claim (and that can be interpreted using 

existing NIOSH dose reconstruction processes or procedures). Dose reconstructions for individuals 

employed at ANL-W during the period from April 10, 1951 through December 31, 1957, but who do 

not qualify for inclusion in the SEC, may be performed using these data as appropriate.
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7.3 Evaluation of Bounding External Radiation Doses at ANL-W 

The principal source of external radiation doses for members of the evaluated class was a combination 

of direct exposures from nuclear reactors, and exposures to the various radionuclide source terms at 

the site, predominately mixed fission products, activation products, and uranium (ORAUT-TKBS0007-6). See Section 5.0 for a more detailed discussion of the radionuclides that are known to have 

been present. 

The following subsections address the ability to bound external doses, methods for bounding doses, 

and the feasibility of external dose reconstruction. 

7.3.1 Evaluation of Bounding Process-Related External Doses

ATTRIBUTION: Section 7.3.1 was completed by Mike Mahathy and Jason Davis, Oak Ridge 

Associated Universities, Brian Gleckler, Dade Moeller, Inc., and Mitch Findley, MJW Corporation. 

All conclusions drawn from the data regarding the specified facilities were peer-reviewed by the 

individuals listed on the cover page. The rationale for all conclusions in subsequent sections of this 

document devoted to the specified facilities are explained in the associated text. 

The following subsections summarize the extent and limitations of information available for 

reconstructing the process-related external doses of members of the class under evaluation.

Personnel Dosimetry Data

NIOSH has captured external monitoring data on ANL-W workers for the entire period of this 

evaluation. However, external monitoring data available from 1952 through 1957 appears to be 

incomplete. NIOSH reviewed external monitoring data supplied for the 50 former workers discussed 

in Section 6.1 for the period 1951 through 1960. Table 7-6 shows the results of this review. NOTE: 

Not all 50 workers were employed during the entire period from 1952 through 1960. Table 7-7 

provides the external monitoring data for 12 workers who were known to be at ANL-W for all years 

from 1952 through 1960, and thus, represents a better assessment of the completeness of records by 

year.

Table 7-6: Available External Monitoring Data for 50 Selected ANL-W Workers

Year Subset of 50 Selected ANL-W Workers with Available External Monitoring Data

1951 18

1952 32

1953 32

1954 32

1955 39

1956 16

1957 18

1958 27

1959 28

1960 20
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External monitoring results available for the same twelve former workers used for Table 7-3 are given 

in Table 7-7 where “partial” denotes that results for only some of the year are available to NIOSH. 

Table 7-7: Available External Monitoring Data for 12 Selected Workers (1952-1960)

Worker 1952 1953 1954 1955 1956 1957 1958 1959 1960

1 Partial complete complete partial none none complete complete complete

2 Partial complete complete partial none none none none none

3 Partial complete complete partial none none complete complete complete

4 Partial complete complete partial partial partial complete complete complete

5 Partial complete complete partial partial partial complete complete complete

6 Partial complete complete partial partial partial complete complete complete

7 Partial complete complete partial partial partial complete complete complete

8 Partial complete complete partial partial partial complete complete complete

9 Partial complete complete partial partial partial complete complete complete

10 Partial complete complete partial partial partial partial complete complete

11 Partial complete complete partial partial partial complete complete complete

12 Partial complete complete partial partial partial complete complete complete

Electron and Photon Dosimetry Data 

Because all workers in ANL-W’s radiological areas were monitored for exposures to photon radiation, 

and because the DOE provides workers’ dosimeter data to NIOSH for dose reconstructions, the 

external electron and photon doses for the monitored workers after 1957 can be reconstructed using 

the guidance in ORAUT-TKBS-0007-6. Because complete sets of dosimetry records are not always 

available for the years prior to 1958, not all of the workers’ doses for the years of 1951–1957 can be 

reconstructed. Because the control dosimeters at ANL-W were likely exposed to elevated levels of 

onsite ambient (environmental) external radiation due to being located at the entrances of the major 

operating areas, onsite ambient external doses also need to be assigned to all monitored ANL-W 

workers to account for any radiation dose that may have been inappropriately subtracted from their 

dosimeter results as background radiation (ORAUT-TKBS-0007-6).

For unmonitored workers, onsite ambient (environmental) external doses may be assigned. The onsite 

ambient external doses are limited to photon doses because the unmonitored workers would not have 

been close enough to the radioactive source terms to receive an external electron dose.

Neutron Dosimetry Data

The majority of the ANL-W workers were not monitored for neutron dose. However, the available 

information indicates that neutron doses were unlikely and that neutron radiation fields are specific to 

a few facilities (ORAUT-TKBS-0007-6). During the period under evaluation, ORAUT-TKBS-0007-

6 indicates that the only location where neutron exposures were possible were at EBR-I, TREAT, 

ZPPR, and HFEF-N’s NRAD Facility (ORAUT-TKBS-0007-6). Recently-captured information, such 

as the references for Section 5.0, indicates that neutron calibration/check sources (e.g., Ra-Be, Po-Be, 

Pu-Be, Am-Be, Cf-252) were present at most of the major facilities at one time or another. Therefore, 
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anyone working near those sources while they were unshielded would have also been exposed to 

neutron radiation. 

Informal observations/reviews of the ANL-W’s neutron dosimetry data, conducted while performing 

dose reconstructions and while capturing records at the ANL-W site, indicate that the neutron doses 

for the monitored workers are typically less than the dosimeters’ limits of detection. An exception 

may have been EBR-I, due to the limitations in the sensitivity of NTA (nuclear track emulsion Type 

A). In the ANL-W claims, career neutron doses of 500 mrem or greater are rare. The available 

information also indicates that ANL-W investigated neutron exposures to unmonitored workers and 

estimated doses for those workers. Given that the ANL-W likely monitored the workers with the 

highest potential to receive neutron doses, and given that the monitored workers’ neutron doses were 

typically less than the dosimeters’ limits of detection, it is unlikely that an unmonitored ANL-W 

worker received more than an incidental exposure to neutron radiation. 

Because unmonitored neutron doses were unlikely at the ANL-W, and because DOE provides the 

workers’ dosimeter data to NIOSH for dose reconstructions, neutron doses for the workers can be 

reconstructed using the guidance in ORAUT-TKBS-0007-6. 

7.3.2 Evaluation of Bounding Ambient Environmental External Doses

ATTRIBUTION: Section 7.3.2 was completed by Mike Mahathy and Jason Davis, Oak Ridge 

Associated Universities, Brian Gleckler, Dade Moeller, Inc., and Mitch Findley, MJW Corporation. 

All conclusions drawn from the data regarding the specified facilities were peer-reviewed by the 

individuals listed on the cover page. The rationale for all conclusions in subsequent sections of this 

document devoted to the specified facilities are explained in the associated text. 

Table 4-13 of ORAUT-TKBS-0007-4 provides the onsite ambient (environmental) external doses 

from 1952 through the end of the period under evaluation. For the year 1951, the onsite ambient 

doses can be bounded by assigning the onsite ambient doses being assigned for the years 1952–1972 

because radiological operations prior to 1952 were minimal compared to 1952 and later (ORAUTTKBS-0007-2).

Because all workers in the ANL-W’s radiological areas were monitored for external dose, and because 

DOE provides workers’ dosimeter data to NIOSH for dose reconstructions, the only doses that need to 

be assigned for unmonitored workers at ANL-W are onsite ambient external doses. Because the 

control dosimeters at ANL-W were likely exposed to elevated levels of onsite ambient radiation due 

to being located at the entrances of the major operating areas, onsite ambient doses are also assigned 

to all monitored ANL-W workers to account for any radiation dose that may have been 

inappropriately subtracted from their dosimeter results as background radiation (ORAUT-TKBS0007-6). 

7.3.3 ANL-W Occupational X-Ray Examinations 

NIOSH has sufficient information through documents in the SRDB and ORAUT-TKBS-0007-3 to 

support the ability to bound the occupational medical doses for the evaluated class. 
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7.3.4 Methods for Bounding External Dose at ANL-W 

There is an established protocol for assessing external exposure when performing dose reconstructions 

(these protocol steps are discussed in the following subsections): 

• Electron and Photon Doses 

• Neutron Dose 

• Occupational Medical Dose

• Ambient Environmental External Dose 

7.3.4.1 Methods for Bounding Operational Period External Dose

Electron and Photon Doses 

While the primary method is the use of individual external dosimetry records provided by DOE, it was 

determined that external doses prior to 1958 could not be adequately reconstructed due to missing 

records. However, dosimeter exchange frequencies and thresholds are known and documented. 

Missed dose may be applied using that information after 1958. The necessary biases/corrections that 

need to be made to the worker’s electron and photon dosimetry results are provided in ORAUTTKBS-0007-6. To account for unmonitored radiation exposures, onsite ambient external doses can be 

assigned in accordance with the guidance in ORAUT-TKBS-0007-4 and ORAUT-PROC-0060. 

Neutron Dose 

While the primary method is the use of individual external dosimetry records provided by DOE, it was 

determined that external doses prior to 1958 could not be adequately reconstructed due to missing 

records. However, dosimeter exchange frequencies and thresholds are known and documented. 

Missed neutron dose may be applied using that information after 1958. The necessary 

biases/corrections that need to be made to the worker’s neutron dosimetry results are provided in 

ORAUT-TKBS-0007-6. Because unmonitored neutron doses were unlikely at ANL-W, unmonitored 

neutron doses do not need to be assigned. 

Occupational Medical Dose 

NIOSH has determined that occupational medical doses can be reconstructed. NIOSH has sufficient 

information in documents in the SRDB and ORAUT-TKBS-0007-3 to support the ability to bound 

occupational medical doses for the evaluated class.

7.3.4.2 Methods for Bounding Ambient Environmental External Doses 

NIOSH has determined that ambient environmental external doses can be reconstructed. The ambient 

environmental external doses can be reconstructed using the guidance in ORAUT-TKBS-0007-4 and 

ORAUT-PROC-0060. 
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7.3.5 External Dose Reconstruction Feasibility Conclusion 

ATTRIBUTION: Section 7.3.5 was completed by Mike Mahathy and Jason Davis, Oak Ridge 

Associated Universities, Brian Gleckler, Dade Moeller, Inc., and Mitch Findley, MJW Corporation. 

All conclusions drawn from the data regarding the specified facilities were peer-reviewed by the 

individuals listed on the cover page. The rationale for all conclusions in subsequent sections of this 

document devoted to the specified facilities are explained in the associated text. 

NIOSH has reached the following conclusions regarding the feasibility of external dose 

reconstruction: 

• External dose for external electron and photon exposures cannot be reconstructed for ANL-W 

workers from April 10, 1951 through December 31, 1957. External dose from electron and 

photon exposures can be reconstructed with sufficient accuracy from January 1, 1958 through 

December 31, 1979. 

• External dose for neutron exposures cannot be reconstructed for ANL-W workers from April 10, 

1951 through December 31, 1957. External dose from neutron exposures can be reconstructed 

with sufficient accuracy from January 1, 1958 through December 31, 1979. 

• Occupational medical dose for occupational X-ray exposures can be reconstructed for all members 

of the evaluated ANL-W class for the entire period under evaluation. 

• Ambient environmental external dose for environmental external exposures can be reconstructed 

for all applicable members of the evaluated ANL-W class during the entire period under 

evaluation. 

Although NIOSH found that it is not possible to completely reconstruct external radiation doses for 

the period from April 10, 1951 through December 31, 1957, NIOSH intends to use any external 

monitoring data that may become available for an individual claim (and that can be interpreted using 

existing NIOSH dose reconstruction processes or procedures). Dose reconstructions for individuals 

employed at ANL-W during the period from April 10, 1951 through December 31, 1957, but who do 

not qualify for inclusion in the SEC, may be performed using these data as appropriate. 
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7.4 Evaluation of Petition Basis for SEC-00224

The following assertion was made on behalf of petition SEC-00224 for ANL-W: 

SEC-00224: There was no internal monitoring for plutonium, neptunium, or fission products at 

ANL-W between 1949 and 1979. 

NIOSH has determined that internal monitoring of personnel at ANL-W focused on fission products 

from the beginning of radiological operations at the site. In-vitro bioassay began in 1952 with gross 

beta and gross beta/gamma analyses and a single uranium analysis. For ANL-W, 1961 marked the 

beginning of in-vivo bioassay in the form of whole-body counting for gamma-emitting radionuclides 

that are common in fission products. In-vitro and in-vivo measurements by year are presented in 

Section 6.1. 

During the course of this evaluation, it was determined that internal monitoring for fission products 

was adequately performed, and that internal monitoring for plutonium and neptunium, while limited, 

was also adequately performed. Based on the information available regarding ANL-W operations, it 

is apparent that Np was only present in relatively small quantities within the irradiated reactor fuels 

and was also present with other actinides and fission products. No bioassay monitoring data specific 

to Np exposures has been identified during the research performed for this ER. Based on the available 

information, neptunium (and other actinides) were never present at the EBR-II Complex without 

mixed fission products also being present. Potential dose from neptunium will be accounted for by

using the actinide ratios in the INL – ANL-W Occupational Internal Dose TBD (ORAUT-TKBS0007-5).

For the years prior to 1958, not all of the external dosimetry and bioassay records are available. Due 

to inadequate availability of the early external dosimetry and bioassay records, NIOSH is 

recommending the period from April 10, 1951 through December 31, 1957 at ANL-W for inclusion in 

the SEC. After 1957, the ANL-W bioassay records being provided by the INL appear to be complete. 

Even though bioassay monitoring for exposures to actinides other than plutonium and neptunium 

(e.g., uranium and thorium) was not adequately performed, those intakes can be estimated using other 

types of monitoring data (e.g., air monitoring data, contamination surveys, etc.). 

7.5 Other Potential SEC Issues Relevant to the Petition Identified During the 

Evaluation

During the feasibility evaluation for SEC-00224, an issue was identified that needs further analysis 

and resolution. The issue and its current status is: 

ISSUE: Verification and validation (V&V) of bioassay data. 

RESPONSE: It is firmly believed that the in-vivo and in-vitro bioassay datasets for ANL-W/INL are 

complete. However, a V&V has not been finished at this time and will be performed as part of the 

development of a future INL internal co-worker study. 
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7.6 Summary of Feasibility Findings for Petition SEC-00224

This report evaluates the feasibility for completing dose reconstructions for employees at the ANL-W 

from April 1951 through December 1979. NIOSH found that the available monitoring records, 

process descriptions and source term data available are not sufficient to complete dose reconstructions 

for the evaluated class of employees.

Table 7-8 summarizes the results of the feasibility findings at ANL-W for each exposure source 

during the time period April 10, 1951 through December 31, 1957 and from January 1, 1958 through 

December 31, 1979. 

Table 7-8: Summary of Feasibility Findings for SEC-00224

April 10, 1951 through December 31, 1957

January 1, 1958 through December 31, 1979

Source of Exposure

April 10, 1951 through 

December 1, 1957 

Reconstruction Feasible 

(Yes or No)

January 1, 1958 through 

December 31, 1979 

Reconstruction Feasible 

(Yes or No)

Internal1 No1 Yes

 - U No Yes

 - Pu No Yes

 - Th No Yes

 - Mixed fission products No Yes

 - Radioiodines No Yes

 - Other transuranic radionuclides No Yes

 - Other radionuclides No Yes

External No2 Yes

 - Gamma No Yes

 - Beta No Yes

 - Neutron No Yes

 - Occupational Medical X-ray Yes Yes

PARTIAL DOSE RECONSTRUCTION INFORMATION: 

1 INTERNAL: For the purpose of partial dose reconstruction for members of the proposed SEC class who worked at 

ANL-W from 1951 through 1957, partial DRs could include dose from any radionuclides for which specific personal 

monitoring data exist that can be assessed by methods defined in approved NIOSH dose reconstruction procedures or 

processes.

2 EXTERNAL: For the purpose of partial dose reconstruction for members of the proposed SEC class who worked at 

ANL-W from 1951 through 1957, partial DRs should include external dose based on the specific personal monitoring data 

that exist that can be assessed by methods defined in approved NIOSH dose reconstruction procedures or processes
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As of January 15, 2016, a total of 273 claims have been submitted to NIOSH for individuals who 

worked at ANL-W during the period under evaluation in this report. Dose reconstructions have been

completed for 237 individuals (~87%). 

Although NIOSH found that it is not possible to completely reconstruct radiation doses for the 

proposed class, NIOSH intends to use any internal and external monitoring data that may become 

available for an individual claim (and that can be interpreted using existing NIOSH dose 

reconstruction processes or procedures). Therefore, dose reconstructions for individuals employed at 

ANL-W during the period from April 10, 1951 through December 31, 1957, but who do not qualify 

for inclusion in the SEC, may be performed using these data as appropriate. 

8.0 Evaluation of Health Endangerment for Petition SEC-00224

The health endangerment determination for the class of employees covered by this evaluation report is 

governed by both EEOICPA and 42 C.F.R. § 83.13(c)(3). Under these requirements, if it is not 

feasible to estimate with sufficient accuracy radiation doses for members of the class, NIOSH must 

also determine that there is a reasonable likelihood that such radiation doses may have endangered the 

health of members of the class. Section 83.13 requires NIOSH to assume that any duration of 

unprotected exposure may have endangered the health of members of a class when it has been

established that the class may have been exposed to radiation during a discrete incident likely to have 

involved levels of exposure similarly high to those occurring during nuclear criticality incidents. If 

the occurrence of such an exceptionally high-level exposure has not been established, then NIOSH is 

required to specify that health was endangered for those workers who were employed for a number of 

work days aggregating at least 250 work days within the parameters established for the class or in 

combination with work days within the parameters established for one or more other classes of 

employees in the SEC. 

Based on the sum of information available from available resources, NIOSH’s evaluation determined 

that it is not feasible to estimate radiation dose with sufficient accuracy for members of the NIOSHevaluated class for the time period from April 10, 1951 through December 31, 1957. NIOSH has 

determined that no ANL-W workers were directly exposed to any of the criticality experiments or 

incidents/accidents that occurred at ANL-W. Therefore, the resulting NIOSH-proposed SEC class 

must include a minimum required employment period of 250 days as a basis for specifying that health 

was endangered for this time period. NIOSH further determined that it is feasible to estimate 

radiation dose with sufficient accuracy for members of the NIOSH-evaluated class for the time period 

from January 1, 1958 through December 31, 1979. Therefore, a health endangerment determination is 

not required for this time period. 
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9.0 Class Conclusion for Petition SEC-00224

Based on its full research of the class under evaluation, NIOSH has defined a single class of 

employees for which NIOSH cannot estimate radiation doses with sufficient accuracy. The NIOSHproposed class to be added to the SEC includes all employees of the Department of Energy, its 

predecessor agencies, and their contractors and subcontractors who worked at the Argonne National 

Laboratory-West between April 10, 1951 and December 31, 1957 for a number of work days 

aggregating at least 250 work days, occurring either solely under this employment, or in combination 

with work days within the parameters established for one or more other classes of employees in the 

Special Exposure Cohort. NIOSH determined that it is feasible to estimate radiation dose with 

sufficient accuracy for members of the NIOSH-evaluated class for the time period from January 1, 

1958 through December 31, 1979. 

NIOSH has carefully reviewed all material sent in by the petitioner, including the specific assertions 

stated in the petition, and has responded herein (see Section 7.4). NIOSH has also reviewed available 

technical resources and many other references, including the Site Research Database (SRDB), for 

information relevant to SEC-00224. In addition, NIOSH reviewed its NOCTS dose reconstruction 

database to identify EEOICPA-related dose reconstructions that might provide information relevant to 

the petition evaluation. 

These actions are based on existing, approved NIOSH processes used in dose reconstruction for 

claims under EEOICPA. NIOSH’s guiding principle in conducting these dose reconstructions is to 

ensure that the assumptions used are fair, consistent, and well-grounded in the best available science. 

Simultaneously, uncertainties in the science and data must be handled to the advantage, rather than to 

the detriment, of the petitioners. When adequate personal dose monitoring information is not 

available, or is very limited, NIOSH may use the highest reasonably possible radiation dose, based on 

reliable science, documented experience, and relevant data to determine the feasibility of 

reconstructing the dose of an SEC petition class. NIOSH contends that it has complied with these 

standards of performance in determining the feasibility or infeasibility of reconstructing dose for the 

class under evaluation. 
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10.0 References 

42 C.F.R. pt. 81, Guidelines for Determining the Probability of Causation Under the Energy 

Employees Occupational Illness Compensation Program Act of 2000; Final Rule, Federal 

Register/Vol. 67, No. 85/Thursday, p. 22,296; May 2, 2002; SRDB Ref ID: 19391 

42 C.F.R. pt. 82, Methods for Radiation Dose Reconstruction Under the Energy Employees 

Occupational Illness Compensation Program Act of 2000; Final Rule; May 2, 2002; SRDB Ref ID: 

19392 

42 C.F.R. pt. 83, Procedures for Designating Classes of Employees as Members of the Special 

Exposure Cohort Under the Energy Employees Occupational Illness Compensation Program Act of 

2000; Final Rule; May 28, 2004; SRDB Ref ID: 22001 

42 U.S.C. §§ 7384-7385 [EEOICPA], Energy Employees Occupational Illness Compensation 

Program Act of 2000, as amended

Abrams, 1975, Reducing Radiation Exposures to as Low as Practicable, memorandum from C.S. 

Abrams to C.A. DeLorenzo; ANL-W-RE-90; Argonne National Laboratory; September 23, 1975; 

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AFSR Quarterly Reports, 1971, Summary of Personnel Exposures; Argonne National Laboratory; 

October 1961-April 1971; SRDB Ref ID: 60294 

Air Sample Data Sheets, 1956, Air Samples at ZPR and EBR; Argonne National Laboratory; February 

21, 1955-June 4, 1956; SRDB Ref ID: 60247 

Air Sample Data Sheets, 1958, Air Sample Data Sheets from EBR and ZPR; Argonne National 

Laboratory; February 3-December 30, 1958; SRDB Ref ID: 60248 

Air Sample Data Sheets, 1959, Air Monitoring, Smear Data, EBR I, ZPR III, AFSR; Argonne National 

Laboratory; April 31, 1959; SRDB Ref ID: 143756, PDF p. 173 

Air Sample Data Sheets, 1963, Borax Air Sample Data and Building Survey; Argonne National 

Laboratory; July 1963; SRDB Ref ID: 145351 

ANL, 2013, Borax - Safety Experiment on a Boiling Water Reactor; Argonne National Laboratory; 

uploaded to YouTube by American Nuclear Society; March 5, 2013; available at: 

http://www.youtube.com/watch?v=8WfNzJVxVz4

ANL, undated, Argonne National Laboratory’s Photos; Flickr stream; accessed 01/13/16; available at:

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excursion/experiment and results; Argonne National Laboratory; October 25, 2006; available at: 

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tml; SRDB Ref ID: 83220, PDF p. 3 
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ANL-PIO, undated, Photographs Relating to Experimental Breeder Reactor II (EBR-II); Argonne 

National Laboratory, Public Information Office; undated; SRDB Ref ID: 87652 

ANL-W S&H Plan, 1992, Site Safety and Health Plan, Argonne National Laboratory West, Analytical 

Laboratory Facility, Argonne National Laboratory-West; August 3, 1992; SRDB Ref ID: 116161 

Annual Report, 1959, Annual Report of the Health and Safety Division 1958; U.S. Atomic Energy 

Commission, Idaho Operations Office; October 15, 1960; SRDB Ref ID: 8079 

Annual Report, 1960, Annual Report of the Health and Safety Division 1959; U.S. Atomic Energy 

Commission, Idaho Operations Office; October 1960; SRDB Ref ID: 8080 

Annual Report, 1961, Annual Report of the Health and Safety Division 1960; U.S. Atomic Energy 

Commission, Idaho Operations Office; September 1961; SRDB Ref ID: 8078 

Annual Report, 1962, Annual Report of the Health and Safety Division 1961; U.S. Atomic Energy 

Commission, Idaho Operations Office; May 21, 1962; SRDB Ref ID: 8076 

Annual Report, 1965, Reactor Physics Division Annual Report: July 1, 1964 to June 30, 1965; ANL7110; Argonne National Laboratory; December 1965; SRDB Ref ID 143300 

Annual Report, 1968, Reactor Physics Division Annual Report: July 1, 1968 to June 30, 1969; 

ANL-7610; Argonne National Laboratory; January 1970; SRDB Ref ID: 136154 

Annual Report, 1969, 1968 Annual Report of the Health Services Laboratory; U.S. Atomic Energy 

Commission, Idaho Operations Office; March 1969; SRDB Ref ID: 1359 

Annual Report, 1970, 1969 Annual Report of the Health Services Laboratory; U.S. Atomic Energy 

Commission, Idaho Operations Office; April 1970; SRDB Ref ID: 1362 

Arave, 1992, Final Report of the Decontamination and Decommissioning of the BORAX-V Facility 

Turbine Building; A.E. Arave, et al.; Idaho National Engineering Laboratory; December 1992, SRDB 

Ref ID: 85313 

Assessment Document, 1980, Zero Power Plutonium Reactor (ZPPR) Building Vault Safety 

Assessment Document (SAD) Vault 64; author unknown; Argonne National Laboratory; September 

1980; SRDB Ref ID: 143882 

Bacca, 1985, Update on the Hot Fuel Examination Facility (HFEF) Complex, J. P. Bacca; Argonne 

National Laboratory-West; October 28, 1985; SRDB Ref ID: 118588 

Bailey, 1958, Fuel Detection Test Borax IV, R.E. Bailey, et al.; Idaho National Laboratory; March 18, 

1958; SRDB Ref ID: 116162 

Batte, 1986, Final Safety Analysis Report for the Fuel Manufacturing Facility Building 704; ANLIFR-57; G.L. Batte; Argonne National Laboratory; December 1986; SRDB Ref ID: 143717 

Beck, 1978, Fast Critical Experiments with Thorium Blankets: Initial Results from ZPPR-8, C. L. 

Beck, et al; Argonne National Laboratory-West; October 25, 1978; SRDB Ref ID: 118384 
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Bess, 2014, Neutron Radiography (NRAD) Reactor 64-Element Core Upgrade, J. D. Bess, et al; Idaho 

National Laboratory; March 2014; SRDB Ref ID: 140507 

Bioassay, 1956-1967, INEL Bioassay CD 2 of 3; author unknown; 1956-1957; SRDB Ref ID: 123096 

Bioassay, 1956-1984, INEL Bioassay CD 1 of 3; author unknown; 1956-1984; SRDB Ref ID: 123095 

Bioassay, 1958-1986, INEL Bioassay CD 3 of 3; author unknown; 1958-1986; SRDB Ref ID: 123104 

Bioassay Instructions, 1975, Instructions for Collecting Bioassay Samples, J. P. Madison; Argonne 

National Laboratory; July 31, 1975; SRDB Ref ID: 143641 

Bioassay Sheets, 1963, ANL Terminations, Idaho Operations Office, Health and Safety Branch; 

Argonne National Laboratory; 1955-63; SRDB Ref ID: 145310 

Black, 1984, Radioactive Liquid Waste Treatment Facility; R.L. Black; Argonne National Laboratory; 

July 1984; SRDB Ref ID: 83769 

Borax HP Documents, 1959, Borax IV Radiation Levels; Argonne National Laboratory; 1959; SRDB 

Ref ID: 138902 

Borax Superheater, 1964, Theoretical Study on the Transfer Function of Borax-V with Central 

Superheater; ANL-6926; J. Wakabayashi; Argonne National Laboratory – Idaho Division; August 

1964; SRDB Ref ID: 147323 

Borax Superheater, 1965, Experiments with Peripheral Superheater Core PSH-1, Borax-V; 

ANL-6962; Argonne National Laboratory – Idaho Division; July 1965; SRDB Ref ID: 147324 

Borax, 1964, Application of Electron-Bombardment Heating for Boiling Liquid Metals; ANL-6869, R. 

E. Holtz; Argonne National Laboratory; March 1964; SRDB Ref ID: 147322 

Borax-IV Test, 1959, Fuel Defect Test — Borax -IV; radiations fields high during operations of March 

11-12, 1958; Argonne National Laboratory; October 1959; SRBD Ref ID: 27049 

Borax-V Description, 1961, Design and Hazards Summary Report Boiling Reactor Experiment V 

(Borax V); ANL-6302; figures 58 and 59 show reactor facilities and control building locations on site 

plan; Argonne National Laboratory; May 1961; SRDB Ref ID: 83142, PDF p. 192 

Borax-V Design and Hazards, 1961, Design and Hazards Summary Report – Boiling Reactor 

Experiment V (BORAX V); ANL-6302; Argonne National Laboratory; May 1961; SRDB Ref ID: 

83142 

Bowman, 1980, Personnel Skin Contamination Report for Laboratory and Offices and HFEF-N, 

memorandum from G.C. Bowman to C.S. Abrams; ANL-W-RE-(HP)-194; Argonne National 

Laboratory; February 19, 1980; SRDB Ref ID: 143639, PDF pp. 77-81 

Brown, 1987, Functional and Operational Design Requirements for Decontamination and 

Decommissioning of the EBR-I Mark-II NaK; B. W. Brown, et al; Idaho National Engineering 

Laboratory; September 28, 1987; SRDB Ref ID: 8179 
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Brunson, 1959, Design and Hazards Report for the Argonne Fast Source Reactor (AFSR); G.S. 

Brunson; Argonne National Laboratory; June 1959; SRDB Ref ID: 1088 

Butler, 1952, Shielding Measurements on EBR-I; letter from J.W. Butler to K. H. Kingdon; Argonne 

National Laboratory; July 23, 1952; SRDB Ref ID: 144996

Carr, 1976, Core-Loading Diagrams for EBR-II Runs 68-81 (December 1973 through February 

1976); A.C. Carr; Argonne National Laboratory; February 1976; SRDB Ref ID: 146141 

Case, 1969, Core-Loading Diagrams for EBR-II Runs 4 Through 38 (April 1965 through midDecember 1969); J.C. Case; Argonne National Laboratory; December 1969; SRDB Ref ID: 148945 

Case, 1971, Core-Loading Diagrams for EBR-II Runs 39 Through 50 (December 1969 through midAugust 1971); J.C. Case; Argonne National Laboratory; September 1971; SRDB Ref ID: 148948 

Case, 1974, Core-loading Diagrams for EBR-II Runs 51-67 (August 1971-December 1973); 

ANL/EBR-079; compiled by J.C. Case; Argonne National Laboratory; January 1974; SRDB Ref ID: 

150141 

Cerutti, 1956, ZPR-III, Argonne’s Fast Critical Facility; B.C. Cerutti, et al.; Nuclear Science and 

Engineering: 1, 126-134 (1956); SRDB Ref ID: 29246 

Cerutti, Jan1956, The Fast Critical Assembly; ANL-5513, B.C. Cerutti, et al.; Argonne National 

Laboratory; January 1956; SRDB Ref ID: 147173 

Cerutti, undated, ZPR-III, Argonne’s New Fast Critical Facility; B. Cerutti, et al.; Argonne National 

Laboratory; undated, SRDB Ref ID: 87647 

Chakraborty, 1971, Performance of EBR-II Driver Fuel Elements from the Start of Reactor Operation 

to the End of Reactor Operation at the 50 MWt Power Level; A.K. Chakraborty; Argonne National 

Laboratory; November 1971; SRDB Ref ID: 146105 

Cipperley, 1959, Improvements in Personnel Metering Procedures at NRTS; F.V. Cipperley, et al.; 

U.S. Atomic Energy Commission, Idaho Operations Office; June 19, 1959; SRDB Ref ID: 8113 

Clean-up Request, 1959, Borax — Dismantling and Cleanup; description of tasks; Argonne National 

Laboratory; July 17, 2959; SRDB Ref ID: 138896, PDF p. 2 

Cold Line Facility, 1969, Preliminary Proposal for Cold-Line Facility, Project No. 69-4-b, 

Modifications to EBR-II; Argonne National Laboratory, National Reactor Testing Station; Idaho Falls, 

Idaho; July 1969; SRDB Ref ID: 145183 

Control Rods, 1957, Operating Instructions for Borax IV; Argonne National Laboratory; November 

1957; SRDB Ref ID: 29234 

Cook, 1969, Operational Health Physics Problems in the EBR-II Fuel Cycle Facility; CONF-690103, 

E. M. Cook, et al; Argonne National Laboratory, Idaho Division; presentation at the 3rd Health 

Physics Society Midyear Topical Symposium, Los Angeles, CA; January 29031, 1969; SRDB Ref ID: 

144226 
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Cook, 1975, Plutonium Contamination Monitoring in the HFEF Hot Cells During Sectioning of 

Mark-IIA Sodium Loops; ANL/HFEF-006; J. H. Cook, et al.; Argonne National Laboratory; May 

1975; SRDB Ref ID: 80575 

Courtney, 1977, Air Sampling System at the Hot Fuel Examination Facility/North; J.C. Courtney, et 

al.; Proceedings of 25th Conference on Remote Systems Technology; pp. 22-27; San Francisco, 

California; November 1977; SRDB Ref ID: 143933 

Courtney, 1991, A Description of the Demonstration Integral Fast Reactor Fuel Cycle Facility; J.C. 

Courtney, et al.; article in Health Physics; 61(4):553-563; 1991; SRDB 131200 

Crewe, 1963, Conceptual Design of the ZPPR Facility; letter from A. V. Crewe (ANL-W) to K. A. 

Dunbar (AEC-Chicago); July 31, 1963; SRDB Ref ID: 145657 

Cushman, 1970, Prospectus: Routine Operation of EBR-II at 62.5 MWt; R.A. Cushman, et al.; 

Argonne National Laboratory; July 1970; SRDB Ref ID: 148946 

Cushman, 1971, Operation of EBR-II as an Irradiation Facility at 62.5 MWt: An Evaluation of Plant 

and Reactor Data from Run 46A; R.A. Cushman, et al.; Argonne National Laboratory; February 1971; 

SRDB Ref ID: 148947 

Davey, 1965, The Ratio of U-238 Capture and U-235 Fission Cross Sections in Fast Reactors; 

ANL-7034; W. G. Davey; Argonne National Laboratory; May 1965; SRDB Ref ID: 147325 

DCAS-PR-004, Internal Procedures for the Evaluation of Special Exposure Cohort Petitions, Rev. 1, 

National Institute for Occupational Safety and Health (NIOSH); Cincinnati, Ohio; April 15, 2011; 

SRDB Ref ID: 94768 

Deitrich, 1998, A Review of Experiments and Results from the Transient Reactor Test (TREAT) 

Facility; ANL/RE/CP-96982; L.W. Deitrich, et al.; presented at American Nuclear Society Winter 

Meeting; Washington, D.C.; November 15-19, 1998; SRDB Ref ID: 128523 

Dickerman, 1979, Summary of TREAT experiments on oxide core-disruptive accidents; ANL-79-13; 

C.E. Dickerman, et al.; Argonne National Laboratory; February 1979; SRDB Ref ID: 150532

Dietrich, 1954, Experimental Investigation of the Self-Limitation of Power During Reactivity 

Transients in a Sub-Cooled, Water-Moderated Reactor, J. R. Dietrich; Argonne National Laboratory; 

1954; SRDB Ref ID: 87986 

Dietrich, Feb1954, Transient and Steady State Characteristics of a Boiling Reactor. The Borax 

Experiments, 1953, AECD-3840; J. R. Dietrich, et al; Argonne National Laboratory; February 1954; 

SRDB Ref ID: 143231 

Disassembly Compliance Report, 1956, Brief of Findings; Argonne National Laboratory; May 23, 

1956; SRDB Ref ID: 145139 

DOE, 1990, Environmental Assessment: Hot Fuel Examination Facility/South; DOE/EA-0377; 

Department of Energy (DOE); Washington, D.C.; May 1990; SRDB Ref ID: 83758 
SEC-00224 02-18-16 ANL-W 

224 of 260 

EBR-1, 1959, Mark IV Plutonium Loading for EBR-I; Argonne National Laboratory; letters from 

various authors and dates in 1958-1959; SRDB Ref ID: 145055 

EBR-I History, 1979, A National Historic Mechanical Engineering Landmark: Experimental Breeder 

Reactor I, Idaho National Engineering Laboratory, The American Society of Mechanical Engineers, 

June 15, 1979; SRBD Ref ID: 83210 

EBR-I Survey Reports, 1963, EBR-I Radiation Survey Reports; Argonne National Laboratory; 

September 1963; SRDB Ref ID: 145142 

EBR-I, undated, Experimental Breeder Reactor-I; summary of history; Idaho National Laboratory; 

undated; SRDB Ref ID: 118372 

EBR-II Fact Sheet and Memo, 1968, Experimental Breeder Reactor No. 2 (EBR-II) Fact Sheet; author 

unknown; Argonne National Laboratory; 1968; SRDB Ref ID: 87654 

Entry Log; 1979; Argon Cell Entry Log June 23, 1978 to October 26, 1979; Argonne National 

Laboratory; October 26, 1979; SRDB Ref ID: 145376 

Environmental Assessment, 1996, Environmental Assessment: Electrometallurgical Treatment 

Research and Demonstration Project in the Fuel Conditioning Facility at Argonne National 

Laboratory – West, DOE/EA-1148; U.S. Department of Energy, Office of Nuclear Energy, Science 

and Technology; May 15, 1996; SRDB Ref ID: 143816, PDF p. 26 

Excuse Codes; unknown; Excuse Codes List; author unknown; Argonne National Laboratory; 

undated; SRDB Ref ID: 150286 

Exposure History, 1976, Exposure History Records; Argonne National Laboratory; 1943-1976; SRDB 

Ref ID: 150294 

Exposure Records, 1972, ZPPR Weekly Exposure Totals; Argonne National Laboratory; 1972; SRDB 

Ref ID: 145507 

Exposure Records, 1973, ZPPR Weekly Exposure Totals; Argonne National Laboratory; 1973; SRDB 

Ref ID: 145512 

Exposure Records, 1974, ZPPR Weekly Exposure Totals; Argonne National Laboratory; 1974; SRDB 

Ref ID: 145454 

Exposure Records, 1975, Radiation Exposure Data, 1971-1975, memorandum from C.S. Abrams to 

C.A. DeLorenzo; Argonne National Laboratory; March 10, 1975; SRDB Ref ID: 150293 

Exposure Records, 1979, Radiation Exposure Data 1971-1978, memorandum from J.P. Madison to 

C.S. Abrams; Argonne National Laboratory; February 6, 1979; SRDB Ref ID: 150283 

Exposure Records, undated A, List of Exposure Records by Box for [Name redacted]; Argonne 

National Laboratory; September 22, 2015; SRDB Ref ID: 150282 
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Exposure Records, undated B, List of Exposure Records by Box for [Name redacted]; Argonne 

National Laboratory; September 21, 2015; SRDB Ref ID: 150284 

Exposure Reports, 1959, Personnel Exposure Report; Argonne National Laboratory; May 20-21, 

1959; SRDB Ref 150292 

Exposure Reports, 1960, Personnel Exposure Report; Argonne National Laboratory; November 15, 

1960; SRDB Ref ID: 150289 

Exposure Resumes, 1966, Dosimetry Branch Activity Reports; Idaho National Laboratory; January 

1966-January 1967; SRDB Ref ID: 125928 

Exposure Resumes, 1970, Dosimetry Branch Activity Reports; Idaho National Laboratory; January 

1967-January 1970; SRDB Ref ID: 125930 

Familiarization Effort, 1970, EBR-II Operating Experience; Atomics International, North American 

Rockwell; September 21, 1970; SRDB Ref ID: 143676 

FASB Surveys, 1983, [FASB] Cold Line Survey Information, FASB routine survey schedule, survey 

maps, survey results, and instrument checks for January 1983; SRDB Ref ID: 143666, PDF p. 35 

FASB Vault, 1972, Fuel Assembly and Storage Building Vault Safety Review, Argonne National 

Laboratory-West; October 26, 1972; SRDB Ref ID: 44396 

FCF Rad Survey, 1970, Radiological Survey Fuel Cycle Facility (FCF), April 1970; Argonne 

National Laboratory; April 1970; SRDB Ref ID: 60316 

FCF Survey Reports, 1963, Radiation Survey Reports, Air Sample Data, Smear Data Sheet, Counter 

Data: Background and Yield Calculations FCF September 1963; Argonne National Laboratory; 

September 1963; SRDB Ref ID: 146515 

FCF Survey Reports, 1965, Radiation Survey Reports, Air Sample Data, Smear Data Sheet, Counter 

Data: Background and Yield Calculations FCF March 1965; Argonne National Laboratory; March 

1965; SRDB Ref ID: 145472 

Film Processing, 1961, EBR-I, Non-Routine Processing, ZPR-III Non-Routine Film Processing; 

Argonne National Laboratory; 1961; SRDB Ref ID: 145064 

Film Reports, 1954, ANL EBR Film Report; Argonne National Laboratory; December 27, 1954; 

SRDB Ref ID: 150288 

Film Reports, 1955, ANL EBR Film Report; Argonne National Laboratory; December 20, 1954 and 

January 4, 1955; SRDB Ref ID: 150287 

Film Reports, 1957, EBR Film Reports; Argonne National Laboratory; December 1956 - January 

1957; SRDB Ref ID: 148838, PDF pp.13-14, 58-61

Findings Report, 1958, Health Physics Activities and Findings on Borax Incident 2-19 to 2-27-58; 

survey documents; Argonne National Laboratory; March 7, 1958; SRDB Ref ID: 116174 
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226 of 260 

Findings Report, 1960, Preliminary Design and Hazards Report — Boiling Reactor Experiment V 

(Borax V); Argonne National Laboratory; describes operation of steam system; February 1960; SRDB 

Ref ID: 137813 

Fish, 1967, Surface Contamination: Proceedings of a Symposium Held at Gatlinburg Tennessee June 

1964; B.R. Fish, ed.; Pergamon Press: Oxford; 1967; SRDB Ref ID: 146886 

Forrester, undated, Safety Aspects of the IFR Pyroprocess Fuel Cycle, CONF-890841-4; R. J. 

Forrester, et al; Argonne National Laboratory; undated - from context, likely late 1980s; SRDB Ref 

ID: 118596, PDF p. 24 

Freund, 1960, Design Summary Report on the Transient Reactor Test Facility Treat; G.A. Freund, et 

al.; Argonne National Laboratory; February 1960; SRDB Ref ID: 1035 

Fuel Cycle Facility, undated, Fuel Cycle Facility (FCF): A Facility for the On-site Processing of 

Spent EBR-II Fuel, 1964-1969; ANL-NN-624; Argonne National Laboratory; SRDB Ref ID: 150060 

Fuel Element Testing, 1958, Borax-IV Reactor: Manufacture of Fuel and Blanket Elements; 

ANL-5721; Argonne National Laboratory; March 1958; SRDB Ref ID: 143308 

Fuel Elements, 1956, Isotopic Assay of EBR Slugs; memorandums about EBR-I fuel elements; 

Argonne National Laboratory; November 13, 1953 to September 11, 1956; SRDB Ref ID: 145048 

Fuel Fabrication, 1957, Manufacture of the ThO2-UO2 Ceramic Fuel Pellets for Borax-IV; Argonne 

National Laboratory; August 1957; SRDB Ref ID: 27051 

Fuel Plate Fabrication, 1958, The Fabrication of Fuel Subassemblies for the Borax III Reactor; 

Argonne National Laboratory; January 1958; SRDB Ref ID: 143291

Graham, 1958, Health Physics Aspects of BORAX IV Run March 11-12, 1958, memorandum from 

E.D. Graham to M. Novick; Idaho National Laboratory; March 20, 1958; SRDB Ref ID: 138306, PDF 

pp. 2-3 

Graham, 1959, The Health Physics Aspects of Borax IV Operation with Defective Fuel Elements; 

purposely inserting a defective fuel element in the core; E.D. Graham and P.G. Stoddart; Argonne 

National Laboratory; 1955-59; SRBD Ref ID: 60252 

Griffiths, 1954, Health Physics Aspects of the Borax Experiment of July 22, 1954, at the National 

Reactor Testing Station; P. Griffiths, et al.; Idaho Operations Office; December 1954; SRDB Ref ID: 

33198 

Haroldsen, 2008, The Story of the Borax Nuclear Reactor and the EBR-I Meltdown, R. Haroldsen; 

2008; SRDB Ref ID: 143311 

Hass, 1979, Fabrication of Driver-Fuel Elements for EBR-II; ANL-79-38; M. Hass, et al.; Argonne 

National Laboratory; September 1979; SRDB Ref ID: 150056 

Health Physics Monthly Report, Jan1968, Health Physics Reports EBR-II 1967; Argonne National 

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Hesson, 1963, Description and Proposed Operation of the Fuel Cycle Facility for the Second 

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HFEF Design, 1976, Structural Design of HFEF/North Operation and Maintenance Manual; HFEF 

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HFEF FSAR, 1977, HFEF/N Neutron Radiography Facility Reactor Final Safety Analysis Report, 

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HFEF History, 1979, History and Status of the Hot Fuel Examination Facility Complex, J. P. Bacca; 

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HFEF Safety Report, 1975, Hot Fuel Examination Facility/North: Facility Safety Report; ANL-7959; 

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HFEF Survey Data, 1972-1973, Air Monitoring, Smear, RWP, Alpha Data for HFEF-N; Argonne 

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Inner Tanks, 1957, Operating Instructions for EBR-I, Mark III; Argonne National Laboratory; 

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IOO Information Packet, 1962, Thumbnail Sketch National Reactor Testing Station; U.S. Atomic 

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ITF Survey Data, 1967-1969, Air Monitoring, Smear, RWP, Alpha Data for ITF and FCF; Argonne 

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Keating, 1975, Appendix B EBR-I Final Radiation Survey Post D&D Operation; S. Keating; Idaho 

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Kendall, 1975, Decontamination and Decommissioning of the EBR-1 Complex Final Report; E.W. 

Kendall, et al.; Idaho National Engineering Laboratory; July 1975; SRDB Ref ID: 83124 

Kirn, 1964, EBR-II Wet Critical Experiments; ANL-6864; F.S. Kirn and W.B. Lowenstein; Argonne 

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Kittel, 1957, The EBR-I Meltdown—Physical and Metallurgical Changes in the Core; J.H. Kittel, et 

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Koch, undated, EBR-II - Experimental Breeder Reactor-II: An Integrated Experimental Fast Reactor 

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L&O Surveys, 1969, L&O Building Radiation Survey Reports for March 1969, selected pages with 

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L&O Surveys, 1973, L&O Building Radiation Survey Reports for June 1973; various authors and 

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Long, 1970, A Documentation of Criticality Data for EBR-II with a Stainless Steel Radial Reflector; 

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Matlock, 1976; Relocation of the AFSR to the ANL East Area Site of the Idaho National Engineering 

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McFarlane, 1976, The High-Pu240 Sector Experiments in ZPPR Assembly 4; H.F. McFarlane, et al.; 

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McFarlane, 1985, ZPPR: Meeting the Reactor Physics Needs of the Nation’s LMR Programs; H.F. 

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Monitoring Data, 1962, Borax Radiation Monitoring Data; Argonne National Laboratory; OctoberDecember 1962; SRDB Ref ID: 145359 

Mueller, 1979, Advanced TREAT Loop – Preliminary Safety Assessment; C.J. Mueller; Argonne 

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Murphy, 1966, Final Disassembly and Examination of the ML-1 Reactor Core; T.L. Murphy, et al.; 

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Neutron Surveys, 1967, Neutron Surveys EBR-II November-January 1967; Argonne National 

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Neutron Survey Sheet, 1966, EBR-I, ZPR-III AFSR Air Sample and Contamination Surveys; Argonne 

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ORAUT-TKBS-0007-4, Technical Basis Document: Idaho National Laboratory and Argonne 

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ORAUT-TKBS-0007-5, Technical Basis Document: Idaho National Laboratory and Argonne 

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ORAUT-TKBS-0007-6, Technical Basis Document: Idaho National Laboratory and Argonne 

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Personal Communication, 2015a, Personal Communication with a former ANL-W chemist; In-person 

Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 18, 2015; SRDB Ref ID: 147021

Personal Communication, 2015b, Personal Communication with a former ANL-W engineer; In-person 

Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 19, 2015; SRDB Ref ID: 147018

Personal Communication, 2015c, Personal Communication with a former ANL-W auditor; In-person 

Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 19, 2015; SRDB Ref ID: 147248

Personal Communication, 2015d, Personal Communication with a former ANL-W operator; In-person 

Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 19, 2015; SRDB Ref ID: 147009

Personal Communication, 2015e, Personal Communication with a former ANL-W manager; In-person 

Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 19, 2015; SRDB Ref ID: 147010

Personal Communication, 2015f, Personal Communication with a current ANL-W operator; In-person 

Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 19, 2015; SRDB Ref ID: 147014
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Personal Communication, 2015g, Personal Communication with a former ANL-W supervisor; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 20, 2015; SRDB Ref ID: 

147015

Personal Communication, 2015h, Personal Communication with a former ANL-W manager; In-person 

Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 20, 2015; SRDB Ref ID: 147016

Personal Communication, 2015i, Personal Communication with a former ANL-W manager; In-person 

Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 20, 2015; SRDB Ref ID: 147017

Personal Communication, 2015j, Personal Communication with a former ANL-W information staffer; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 20, 2015; SRDB Ref ID: 

147019

Personal Communication, 2015k, Personal Communication with a former ANL-W manager; In-person 

Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 20, 2015; SRDB Ref ID: 147020

Personal Communication, 2015L, Personal Communication with a former ANL-W trainer; In-person 

Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 21, 2015; SRDB Ref ID: 147011

Personal Communication, 2015m, Personal Communication with a current ANL-W operator; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 21, 2015; SRDB Ref ID: 

147012

Personal Communication, 2015n, Personal Communication with a former ANL-W engineer; In-person 

Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 21, 2015; SRDB Ref ID: 147013

Personal Communication, 2015o, Personal Communication with a former ANL-W pipefitter; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; May 21, 2015; SRDB Ref ID: 

147022

Personal Communication, 2015p, Personal Communication with a former ANL-W supervisor; 

In-person Interview by ORAU Team, NIOSH, ABRWH, and SC&A; July 22, 2015; SRDB Ref ID: 

147028

Personal Communication, 2015q, Personal Communication with a former ANL-W manager; In-person 

Interview by NIOSH and ABRWH; July 22, 2015; SRDB Ref ID: 147249

Personal Communication, 2015r, Personal Communication with a former ANL-W technician

In-person Interview by [ORAU Team or NIOSH]; July 22, 2015; SRDB Ref ID: 147446

Personal Communication, 2015s, Personal Communication with a former ANL-W designer; In-person 

Interview by ORAU Team, NIOSH, ABRWH, and SC&A; July 22, 2015; SRDB Ref ID: 147445 

Personal Communication, 2015t, Personal Communication with a former ANL-W reactor operator; 

In-person Interview by ORAU Team and NIOSH; November 4, 2015; SRDB Ref ID: 150163 
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Personnel Exposure Reports, 1962, Report of Radiation Exposures Resulting in the Internal Body 

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Personnel Metering, 1958, Personnel Metering Operating Techniques Self-Service System; U.S. 

Atomic Energy Commission, Idaho Operations Office; 1951-1958; SRDB Ref ID: 8112 

Personnel Metering Tables, 1962, Personnel Metering Branch Activity Reports; Idaho National 

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Personnel Metering Tables, 1963, Personnel Metering Branch Activity Reports; Idaho National 

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Personnel Metering Tables, 1964, Dosimetry Branch Activity Reports; Idaho National Laboratory; 

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Personnel Surveys, 1970, Health Physics Report; Argonne National Laboratory; November 12, 1969-

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Peterson, 2004, Supplement to Technical Basis Document 4 for the Idaho National Engineering and 

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Plutonium Loading, 1965, ZPR-III Loading for SEFOR Critical Experiments; ANL-7049; Idaho 

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Policy Letters, 1966, Radiation Protection Testing and Procedures; Idaho National Nuclear 

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Procedures, 1963, Health and Safety Branch Standard Practice Manual; Phillips Petroleum Company, 

Atomic Energy Division, Idaho Falls, Idaho; whole body counting procedures; April 2, 1963; SRDB 

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Progress Report, 1953, Progress Report on the Experimental Breeder Reactor: April 1, 1951 Through 

January 31, 1953; ANL-5023; Argonne National Laboratory; February 20, 1953; SRBD Ref ID: 

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Progress Report, 1962, Reactor Development Program Progress Report: September 1962; ANL-6619; 

Argonne National Laboratory; October 15, 1962; SRDB Ref ID: 143239 

Progress Report, Mar1963, Reactor Development Program Progress Report: March 1963; 

ANL-6705; Argonne National Laboratory; April 15, 1963; SRDB Ref ID: 143248 

Progress Report, Aug1963, Reactor Development Program Progress Report: August 1963; 

ANL-6780; Argonne National Laboratory; September 13, 1963; SRDB Ref ID: 143254 

Progress Report, Nov1963, Reactor Development Program Progress Report: November 1963; 

ANL-6808; Argonne National Laboratory; December 15, 1963; SRDB Ref ID: 143257 
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Progress Report, Dec1963, Reactor Development Program Progress Report: December 1963; 

ANL-6810; Argonne National Laboratory; December, 1963; SRDB Ref ID: 143259 

Progress Report, Jan1964, Reactor Development Program Progress Report: January 1964; 

ANL-6840; Argonne National Laboratory; February 15, 1964; SRDB Ref ID: 143262 

Progress Report, Feb1964, Reactor Development Program Progress Report: February 1964; 

ANL-6860; Argonne National Laboratory; March 15, 1964; SRDB Ref ID: 143263 

Progress Report, Mar1964, Reactor Development Program Progress Report: March 1964; 

ANL-6880; Argonne National Laboratory; April 15, 1964; SRDB Ref ID: 143264 

Progress Report, Apr1964, Reactor Development Program Progress Report: April 1964; ANL-6885; 

Argonne National Laboratory; May 15, 1964; SRDB Ref ID: 143266 

Progress Report, May1964, Reactor Development Program Progress Report: May 1964; ANL-6904; 

Argonne National Laboratory; June 15, 1964; SRDB Ref ID: 143268 

Progress Report, Jun1964, Reactor Development Program Progress Report: June 1964; ANL-6912; 

Argonne National Laboratory; July 15, 1964; SRDB Ref ID: 143269 

Progress Report, Jul1964, Reactor Development Program Progress Report: July 1964; ANL-6923; 

Argonne National Laboratory; August 20, 1964; SRDB Ref ID: 143271 

Progress Report, Aug1964, Reactor Development Program Progress Report: August 1964; 

ANL-6936; Argonne National Laboratory; August 1964; SRDB Ref ID: 143272 

Progress Report, Sep1964, Reactor Development Program Progress Report: September 1964; 

ANL-6944; Argonne National Laboratory; September 1964; SRDB Ref ID: 143274 

Progress Report, Jan1965, Reactor Development Program Progress Report: January 1965; 

ANL-7003; Argonne National Laboratory; January 1965; SRDB Ref ID: 143281 

Progress Report, Mar1965, Reactor Development Program Progress Report: March 1965; 

ANL-7024; Argonne National Laboratory; April 20, 1965; SRDB Ref ID: 143282 

Progress Report, Jun1965, Reactor Development Program Progress Report: June 1965; ANL-7071; 

Argonne National Laboratory; July 30, 1965; SRDB Ref ID: 143287 

Progress Report, Dec1965, Reactor Development Program Progress Report: December 1965; ANL7132; Argonne National Laboratory; January 25, 1966; SRDB Ref ID: 150114 

Progress Report, 1966, Reactor Development Program Progress Report: May 1966; ANL-7219; 

Argonne National Laboratory; June 20, 1966; SRDB Ref ID: 137812 

Progress Report, Apr1967, Reactor Development Program Progress Report: April 1967; ANL-7329; 

Argonne National Laboratory; June 16, 1967; SRDB Ref ID: 150122 
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Progress Report, May1967, Reactor Development Program Progress Report: May 1967; ANL-7342; 

Argonne National Laboratory; June 30, 1967; SRDB Ref ID: 150123 

Progress Report, Jun1967, Reactor Development Program Progress Report: June 1967; ANL-7349; 

Argonne National Laboratory; July 31, 1967; SRDB Ref ID: 136152 

Progress Report, Jul1967, Reactor Development Program Progress Report: July 1967; ANL-7357; 

Argonne National Laboratory; August 31, 1967; SRDB Ref ID: 150126 

Progress Report, Aug1967, Reactor Development Program Progress Report: August 1967; ANL7371; Argonne National Laboratory; September 28, 1967; SRDB Ref ID: 136221 

Progress Report, Sept1967, Reactor Development Program Progress Report: September 1967; ANL7382; Argonne National Laboratory; December 31, 1967; SRDB Ref ID: 150127 

Progress Report, Nov1967, Reactor Development Program Progress Report: November 1967; 

ANL-7399; Argonne National Laboratory; November 1967; SRDB Ref ID: 147326 

Progress Report, Dec1967, Reactor Development Program Progress Report: December 1967; ANL7403; Argonne National Laboratory; January 1, 1968; SRDB Ref ID: 136147 

Progress Report, Jan1968, Reactor Development Program Progress Report: January 1968; 

ANL-7419; Argonne National Laboratory; February 27, 1968; SRDB Ref ID: 147255 

Progress Report, Feb1968, Reactor Development Program Progress Report: February 1968; 

ANL-7427; Argonne National Laboratory; March 26, 1968; SRDB Ref ID: 147256 

Progress Report, Apr1968, Reactor Development Program Progress Report: April 1968; ANL-7445; 

Argonne National Laboratory; June 12, 1968; SRDB Ref ID: 118614 

Progress Report, May1968, Reactor Development Program Progress Report: May 1968; ANL-7457; 

Argonne National Laboratory; July 2, 1968; SRDB Ref ID: 150129 

Progress Report, Oct1968, Reactor Development Program Progress Report: October 1968; ANL7513; Argonne National Laboratory; November 27, 1968; SRDB Ref ID: 150133 

Progress Report, Dec1968, Reactor Development Program Progress Report: December 1968; ANL7527; Argonne National Laboratory; January 29, 1969; SRDB Ref ID: 148939 

Progress Report. Jan1969, Reactor Development Program Progress Report: January 1969; ANL7548; Argonne National Laboratory; February 28, 1969; SRDB Ref ID: 148941 

Progress Report, Mar1969, Reactor Development Program Progress Report: March 1969; ANL7561; Argonne National Laboratory; April 22, 1969; SRDB Ref ID: 148943 

Progress Report, Feb1969, Reactor Development Program Progress Report: February 1969; 

ANL-7553; Argonne National Laboratory; March 25, 1969; SRDB Ref ID: 148942 
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Progress Report, Dec1969, Reactor Development Program Progress Report: December 1969; ANL7655; Argonne National Laboratory; January 29, 1970; SRDB Ref ID: 150139 

Progress Report, Jan1970, Reactor Development Program Progress Report: January 1970; ANL7661; Argonne National Laboratory; February 25, 1970; SRDB Ref ID: 147382 

Progress Report, Jul1970, Reactor Development Program Progress Report: July 1970; ANL-7726; 

Argonne National Laboratory; August 27, 1970; SRDB Ref ID: 147384 

Progress Report, Aug1970, Reactor Development Program Progress Report: August 1970; 

ANL-7737; Argonne National Laboratory; September 29, 1970; SRDB Ref ID: 147385 

Progress Report, Mar1971, Reactor Development Program Progress Report: March 1971; 

ANL-7798; Argonne National Laboratory; April 19, 1971; SRDB Ref ID: 148944 

Progress Report, AprMay1971, Reactor Development Program Progress Report: April-May 1971; 

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Puphal, 1996, INEL External Dosimetry History; K. Puphal; Idaho National Laboratory; June 10, 

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Radiation Fields, 2011, Idaho National Laboratory and Argonne National Laboratory-West—

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Radionuclide Concentrations, undated, Location and Radionuclide Concentration of Near Surface 

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Reactor Operation, 1962, Idaho Division Summary Report; Argonne National Laboratory; October 

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Reactor Quarterly, 1954, Reactor Engineering Division Quarterly Report: July 1, 1954 through 

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Reactor Quarterly, 1955, Quarterly Progress Report on Reactor Development January 31, 1955; 

ANL-5398; Argonne National Laboratory; January 31, 1955; SRDB Ref ID: 143253 

Reactor Quarterly, 1956a Reactor Engineering Division Quarterly Report: January, February, 

March, 1956; ANL-5571; Argonne National Laboratory; July 1956; SRDB Ref ID: 143301 

Reactor Quarterly, 1956b, Quarterly Progress Report on Reactor Development: 4000 Program; ANL5594; Argonne National Laboratory; April 30, 1956; SRDB Ref ID: 143303 
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Reactor Steady-State Operation, 1956; Operational Experience with Borax Power Plant; identify 

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Record Sheets, 1958-1965, IDO H and S Sample Record Sheets, EBR-I and EBR-II; Argonne National 

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Record Sheets, 1959-1965, IDO H and S Sample Record Sheets, EBR-I and EBR-II; Argonne National 

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Record Sheets, 1962, IDO H and S Sample Record Sheets and Personnel Exposure History; Argonne 

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Record Sheets, Apr-May1962, IDO H and S Sample Record Sheets for Burial Ground; Argonne 

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Record Sheets, 1963, IDO H and S Sample Record Sheets for Thomas William Asbury; Argonne 

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Record Sheets, 1963-1965, IDO H and S Sample Record Sheets, Daily Personnel Metering Log, 

Previous Occupational Radiation Exposure Form, and Whole-Body Gamma-Ray Analysis Form; 

Argonne National Laboratory; 1963-1965; SRDB Ref ID: 150295 

Record Sheets, 1964, IDO H and S Sample Record Sheet, EBR-II; Argonne National Laboratory; 

November 6, 1964; SRDB Ref ID: 150300 

Record Sheets, 1966, IDO H and S Sample Record Sheets, EBR-I and EBR-II; Argonne National 

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Request for Pellets, 1957, Borax IV Type Thoria Urania Pellets; experiment to simulate a fuel rupture; 

Argonne National Laboratory; Nov. 11, 1957; SRBD Ref ID: 116176, PDF p. 2 

Research Highlights, 1965, Chemical Engineering Division Research Highlights May 1964 - April 

1965; ANL-7020; Argonne National Laboratory; 1965; SRDB Ref ID: 150108 

Richards, 1979, Neutron Radiography at the Hot Fuel Examination Facility, W. J. Richards, et al; 

Proceedings of 27th Conference on Remote Systems Technology, San Francisco, November1979; 

Argonne National Laboratory-West; November 1979; SRDB Ref ID: 1062 

Robinson, 1971, The Mark II Integral Sodium TREAT Loop; ANL-7692; L.E. Robinson, et al.; 

Argonne National Laboratory; November 1971; SRDB Ref ID: 150057 

Sackett, 2008, EBR-II Test and Operating Experience, J. I. Sackett, Idaho National Laboratory; 

December 5, 2008; SRDB Ref ID: 143817 

Sackett, undated, Operating and Test Experience of EBR-II, J.I. Sackett; Argonne National 

Laboratory; undated; SRDB Ref ID: 118590 
SEC-00224 02-18-16 ANL-W 

238 of 260 

Safety Monthly Report, 1956, Industrial Hygiene and Safety Division Radiation Safety Section 

Monthly Report; Argonne National Laboratory; February-June 1956; SRDB Ref ID: 145767 

Safety Monthly Report, 1957, Industrial Hygiene and Safety Division Radiation Safety Section 

Monthly Report; Argonne National Laboratory; July-December 1957; SRDB Ref ID: 145807 

Safety Monthly Report, 1958, Industrial Hygiene and Safety Division Radiation Safety Section 

Monthly Report; Argonne National Laboratory; January-December 1958; SRDB Ref ID: 145799 

Safety Monthly Report, 1959, Industrial Hygiene and Safety Division Radiation Safety Section 

Monthly Report; Argonne National Laboratory; January 1959; SRDB Ref ID: 145775, PDF p. 145 

Safety Monthly Report, 1960, Industrial Hygiene and Safety Division Radiation Safety Section 

Monthly Report; Argonne National Laboratory; May-January 1960; SRDB Ref ID: 145771 

Safety Monthly Report, 1961, Industrial Hygiene and Safety Division Radiation Safety Section 

Monthly Report; Argonne National Laboratory; December 1961; SRDB Ref ID: 145816; PDF p. 217 

Safety Monthly Report, Jul1962, Industrial Hygiene and Safety Division Radiation Safety Section 

Monthly Report; Argonne National Laboratory; July 1962; SRDB Ref ID: 145825; PDF p. 122 

Safety Monthly Report, Aug1962, Industrial Hygiene and Safety Division Radiation Safety Section 

Monthly Report; Argonne National Laboratory; August 1962; SRDB Ref ID: 145825; PDF p. 95 

Safety Monthly Report, 1963, Industrial Hygiene and Safety Division Radiation Safety Section 

Monthly Report; Argonne National Laboratory; January-November 1963; SRDB Ref ID: 145675 

Safety Report, 1983, Radioactive Scrap and Waste Facility (RSWF) Safety Analysis Report (SAR); 

Argonne National Laboratory; March 1983; SRDB Ref ID: 143897 

Safety Services, undated, Special Safety Features of EBR-I and Borax-III Reactors at Idaho Division; 

Argonne National Laboratory; undated; SRDB Ref ID:138326 

Sample Data Sheets, 1955, EBR-I and ZPRIII Air and Water Samples; Argonne National Laboratory; 

March 24-December 29, 1955; SRDB Ref ID: 60246 

Sample Data Sheets, 1961, Smear and Air Sample Data Sheets from EBR-I; Argonne National 

Laboratory; November 4, 1961-December 29, 1961; SRDB Ref ID: 138352 

Sehlke, 2010, Idaho National Laboratory Materials and Fuels Complex Natural Phenomena Hazards 

Flood Assessment, INL/EXT-10-20572; G. Sehlke and P. Wichlacz; Idaho National Laboratory; 

December 2010; SRDB Ref ID: 150598, graphic derived from figure on PDF p. 14 

Seidel, 1986, Experience with EBR-II Driver Fuel; B.R. Seidel, et al.; Argonne National Laboratory; 

presented at International Conference on Reliable Fuels for Liquid Metal Rectors, Tucson, Arizona, 

September 7-11, 1986; SRDB Ref ID: 149104 
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239 of 260 

Severn, 1969, Trip Settings on All Particulate Air Monitors in the ZPPR Facility, memorandum from 

F. Severn to P.I. Amundson; Argonne National Laboratory-West; January 31, 1969; SRDB Ref ID: 

143634 

Severn, 1975, Weekly Report 1/20/75 through 1/26/75 – ZPPR-AFSR; memorandum from F. W. 

Severn to J.P. Madison; Argonne National Laboratory; January 28, 1975; SRDB Ref ID: 146479 

Severn, 1977, Radiation Safety and ZPPR-AFSR Reports, ; memoranda from F. W. Severn to C. E. 

Holson; Argonne National Laboratory; December 27, 1976-January 1, 1978; SRDB Ref ID: 145538 

Shielding Enhancements, 1960, Idaho Division Summary Report; Argonne National Laboratory; JulySeptember 1960; SRDB Ref ID: 118563, PDF p. 148 

Shipping Papers, 1976, Shipping Papers for Enriched Uranium from SRS to NLO; National Lead 

Company of Ohio; 1976-1977; SRDB Ref ID: 43767 

Site Plan, 1981, Master Site Plan: ANL-W; ANL-NN42; Argonne National Laboratory-West; 

February 1981; SRDB Ref ID: 146142 

Smith, 1965, Terminal Report for the Mark IV (Plutonium) Loading in EBR-I; ANL-6865, R. R. 

Smith, et al; Argonne National Laboratory; April 1965; SRDB Ref ID: 138877 

Smith, 1977, EBR-II Operating Experience; CONF-770822-6; R.N. Smith, et al.; Argonne National 

Laboratory; prepared for 8th Conference on Reactor Operating Experience; Chattanooga, TN; August 

8-10, 1977; SRDB Ref ID: 150032 

Smith, 1980, Borax-I Radiological Characterization; D.L. Smith; July 1980; Waste Management 

Program EG&G, Inc.; SRDB Ref ID: 83007 

Smith, 2001, Spent Fuel Fissile Mass Verification at the ANL-West Radioactive Scrap and Waste 

Facility; L.E. Smith, et al.; Argonne National Laboratory; presentation at WM’01 Conference; 

Tucson, Arizona; February 25-March 1, 2001; SRDB Ref ID: 83815 

Stacy, 2000, Proving the Principle; Susan M. Stacy; U.S. Department of Energy, Idaho Operations 

Office; 2000; SRDB Ref ID: 1152 

Staker, 1975, The Use of EBR-II as a Principal Fast Breeder Reactor Irradiation Test Facility in the 

U.S.; R.G. Staker, et al.; Argonne National Laboratory; June 1975; SRDB Ref ID:118591 

Status Summaries - Borax IV, 1958, Status Summaries - Borax IV, Idaho National Laboratory; 

January-May 1958, SRDB Ref ID: 138302 

Stephens, 1976; Argonne Facilities: Liquid Metal Fast Breeder Reactor Program; ANL-LEO-TR-76-

2; S.V. Stephens; Argonne National Laboratory; September, 1976; SRDB Ref ID: 118620 

Stevenson, 1969, Operation of a Pyrochemical Processing-Remote Re-fabrication Plant: The EBR-II 

Fuel Cycle Facility; C.E. Stephenson, et al.; Chemical Engineering Progress 65: Symposium Series 

No. 94, pp. 61-9; 60th Annual Meeting of the American Institute of Chemical Engineers; New York, 

New York; 1967; SRDB Ref ID: 144688
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Stevenson, 1987, The EBR-II Fuel Cycle Story, C. E. Stevenson; American Nuclear Society; La 

Grange, IL; 1987; SRDB Ref ID: 144232 

Stoddart, 1963, The Plutonium Aerosol Monitoring Program at ANL-Idaho Facilities; ANL-6759; P. 

G. Stoddart; Argonne National Laboratory; July 1963; SRDB Ref ID: 136529 

Stoddart, 1967, Health Physics Aspects of Cold Line Operations: Initial Progress Report, 

memorandum from P.G. Stoddart to E.D. Graham; Argonne National Laboratory; Idaho Falls, Idaho; 

August 29, 1967; SRDB Ref ID: 145181 

Summary Report, 1953, Argonne National Laboratory, Chemical Engineering Division, Summary 

Report: January, February, and March 1953; ANL-5039; Argonne National Laboratory, Lemont, 

Illinois; May 12, 1953; SRDB Ref ID: 138049 

Summary Report, 1960, Argonne National Laboratory, Idaho Division, Summary Report: July, 

August, September, 1960; ANL-6301; Argonne National Laboratory; 1960; SRDB Ref ID: 118563 

Summary Report, 1961, Argonne National Laboratory, Idaho Division, Summary Report: April 

through September 1961; ANL-6539; Argonne National Laboratory; September 1961; SRDB Ref ID: 

143286 

Survey Reports, Mar1962; EBR-I Radiation Survey Report; Argonne National Laboratory; March 

1962; SRDB Ref ID: 60250 

Survey Reports, May1962, Borax-V Radiation Survey Reports; Argonne National Laboratory; May 

1962; SRDB Ref ID: 145207 

Survey Reports, May-Aug1962; Air Monitoring, Smears, Borax V, EBR-I Neutron; Argonne National 

Laboratory; May-August 1962; SRDB Ref ID: 143762 

Survey Reports, Aug1962, Borax Radiation Survey Reports; Argonne National Laboratory; August 

1962; SRDB Ref ID: 145215 

Survey Reports, 1963, Examples of Radiation Surveys; Argonne National Laboratory; 1963; SRDB 

Ref ID: 145371 

Survey Reports, Jan-Feb1963, EBR-I Radiation Survey Reports; Argonne National Laboratory; 

January-February 1963; SRDB Ref ID: 145240 

Survey Reports, Feb1963, Borax Air Sample and Building Survey; Argonne National Laboratory; 

February 1963; SRDB Ref ID: 145386 

Survey Reports, Mar1963, EBR-I Radiation Survey Reports; Argonne National Laboratory; March 

1963; SRDB Ref ID: 145587, PDF p. 2 

Survey Reports, Jun1963, Radiation Survey Reports and Air Sample Data for Borax; Idaho National 

Laboratory; June 1963; SRDB Ref ID: 138148 
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Survey Reports, Jun-Aug1963, Smear, Air Monitoring Data, Borax-V; Argonne National Laboratory; 

June-August 1963; SRDB Ref ID: 145523 

Survey Reports, Jul1963; EBR-I Radiation Survey Reports; Argonne National Laboratory; July 1-22, 

1963; SRDB Ref ID: 145585 

Survey Reports, Mar1964, ZPR-III Radiation Survey Reports; Argonne National Laboratory; March 

22, 1964; SRDB Ref ID: 145685

Survey Reports, Apr1964, Air Sample Data, Radiation Survey Reports, Smear Data Sheets and Smear 

Sample Data EBR-II and FCF April 1964; Argonne National Laboratory; April 1964; SRDB Ref ID: 

145363 

Survey Reports, Jan1965, EBR-I, ZPR-III, AFSR Radiation Survey Reports; Argonne National 

Laboratory; January-May 1965; SRDB Ref ID: 145556 

Survey Reports, Jun1965, West Area Radiation Survey Reports; Argonne National Laboratory; JuneSeptember 1965; SRDB Ref ID: 145558 

Survey Reports, Dec1965, EBR-II Fuel Cycle Facility Radiation Survey Reports; Argonne National 

Laboratory; December 1965; SRDB Ref ID: 145755 

Survey Reports, Feb1969, Radiation Survey Reports, Smear Sample, Air Sample Data EBR-I, FCF, 

ITF, L and O February - May 1969; Argonne National Laboratory; February-May 1969; SRDB Ref 

ID: 145093 

Survey Reports, July1969, Radiation Survey Reports, Smear Sample Data and Smear Sample Sheets 

FCF July 1969; Argonne National Laboratory; July 1969; SRDB Ref ID: 145407

Survey Reports, Aug-Dec1969, Radiation Surveys and Air Samples, Counter Data, EBR-II August - 

December 1969; Argonne National Laboratory; August-December 1969; SRDB Ref ID: 143709 

Survey Reports, Nov1969, Radiation Survey Reports, Smear Sample Data, Smear Data Sheets and Air 

Sample Data FCF November 1969; Argonne National Laboratory; November 1969; SRDB Ref ID: 

145251 

Survey Reports, Feb1970, Radiation Survey Reports, Smear Sample Data, Smear Data Sheets and Air 

Sample Data FCF February 1970; Argonne National Laboratory; February 1970; SRDB Ref ID: 

145242

Survey Reports, Apr1970, Fuel Cycle Facility Radiation Survey Reports; Argonne National 

Laboratory; April 1970; SRDB Ref ID: 145237 

Survey Reports, Aug1970, Radiation Survey Reports, Smear Sample Data and Air Sample Data FCF 

August 1970; Argonne National Laboratory; August-December 1970; SRDB Ref ID: 145231

Survey Reports, Sept-Dec1970, Radiation Survey Reports, Air Sample Data, Smear Sample Data, 

Smear Data Sheets and Counter Data: Background and Yield Calculations ZPPR SeptemberDecember 1970; Argonne National Laboratory; September-December 1970; SRDB Ref ID: 146015 
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Survey Reports, Oct1970, Radiation Survey Reports, Smear Sample Data, Smear Data Sheets and Air 

Sample Data FEF October 1970; Argonne National Laboratory; October 1970; SRDB Ref ID: 

145233 

Survey Reports, 1972, Radiation Survey Reports, TREAT Radiography, Transient Data Sheets and Air 

Sample Data TREAT Reactor April 1972; Argonne National Laboratory; April 1972; SRDB Ref ID: 

145526 

Survey Reports, 1973, Air Radiation Survey Reports, Radiation Information Reports, Smear Sample 

Data, Smear Sample Sheets and Air Samples Data HFEF-S January - April 1973; Argonne National 

Laboratory; April 1973; SRDB Ref ID: 145594 

Survey Reports, 1976, RWPs, Rad Surveys, Air Sampling Data HFEFIN July-December 1976; 

Argonne National Laboratory; July-December 1976; SRDB Ref ID: 145723

Survey Reports, 1978, Air Sample Data, Survey Maps, Work Permit, FASB, ITF, RMMS July 1978; 

Argonne National Laboratory; April 1970; SRDB Ref ID: 143678 

Till and Chang, 2011, Plentiful Energy: The Story of the Integral Fast Reactor. The Complex History 

of a Simple Reactor Technology, with Emphasis on its Scientific Basis for Non-specialists; C.E. Till 

and Y.I. Chang; Argonne National Laboratory; 2011; SRDB Ref ID: 130385 

TREAT Facility, 1971, Description of TREAT Facility (Revised Section III of Hazards Summary 

Report on the Transient Reactor Test Facility (TREAT), Addendum 3; ANL-5923; Argonne National 

Laboratory; September, 1971; SRDB Ref ID: 143898 

TREAT Hazards Summary, 1958, Hazards Summary Report on the Transient Reactor Test Facility 

(TREAT); ANL-5923; Argonne National Laboratory; October 1958; SRDB Ref ID: 85324 

TREAT Manual, 1971, Transient Reactor Test Facility (TREAT), operating manual, Argonne National 

Laboratory-West; September 1971; SRDB Ref ID: 145613 

Uranium Core Configurations, 1961, Design and Hazards Summary Report Boiling Reactor 

Experiment V (Borax V); ANL-6302; design of three core configurations and coolant; Argonne 

National Laboratory; May 1961; SRDB Ref ID: 83142, PDF p. 16-14 

Walters, 1999, Thirty Years of Fuels and Materials Information from EBR-II; L. C. Walters; Argonne 

National Laboratory; article in Journal of Nuclear Materials 270:39-48; 1999; SRDB Ref ID: 149105 

Waste Discharges, 1958, ICPP Monthly Radioactive Waste Report December 1-31, 1957; Phillips 

Petroleum Company; January 7, 1958; SRDB Ref ID: 139175 

Water Pumps, 1956, Operational Experience with Borax Power Plant; description of steam pressure 

and cooling tower operations; Argonne National Laboratory; June 18, 1956; SRDB Ref ID: 29242, 

PDF p. 4-5 

Wilson, 1961, Film Capabilities for High Level Dose Evaluation; R.H. Wilson, et al.; Hanford 

Laboratories Operation, General Electric Company; September 18, 1961, SRDB Ref ID: 62590 
SEC-00224 02-18-16 ANL-W 

243 of 260 

Witbeck, 1976, Description of the RSWF, the Liners, Cans and Material; ANL-W-RE-(ESE)-177; 

letter from L.C. Witbeck to V. F. Boston; Argonne National Laboratory; December 7, 1976; SRDB 

Ref ID: 143879 

Zinn, 1956, Operational Experience with Borax Power Plant; description of reactor and system; 

Argonne National Laboratory; article in Nuclear Science and Engineering; 1:420-437; June 18, 1956; 

SRDB Ref ID: 29242, PDF p. 2 

Zinn, Jun1956, A Letter on EBR-I Fuel Meltdown; W.H. Zinn; Argonne National Laboratory; article 

in Nucleonics; 14(6):35; June 1956; SRDB Ref ID: 27052 

ZPPR, 1969, Technical Specifications for the Zero Power Plutonium Reactor; Argonne National 

Laboratory-West; July 10, 1969; SRDB Ref ID: 138139 

ZPPR, 1970, Reactor Safety Survey Report: Zero Power Plutonium Reactor (ZPPR); Argonne 

National Laboratory; 1970; SRDB Ref ID: 140237 

ZPPR, 1984, Loading and Unloading Procedures for Zero Power Plutonium Reactor (ZPPR); 

Argonne National Laboratory; February, 1984; SRDB Ref ID: 144681 

ZPPR, 1973, ZPPR Radiological Surveys for February 1973, Argonne National Laboratory-West; 

various dates in February 1973; SRDB Ref ID: 60328 

ZPPR, undated, Cross Section Diagram ZPPR; Argonne National Laboratory; undated; SRDB Ref 

ID: 143597 

ZPPR Dosimetry Policy, 1971, Dosimetry Branch Activity Report for October 21 through November 

20, 1971; Idaho National Laboratory; November 30, 1971; SRDB Ref ID: 125934, PDF p. 8 

ZPPR Surveys, 1976, Zero Power Plutonium Reactor (ZPPR) Routine Radiological Surveys, Argonne 

National Laboratory-West; various authors and dates in August 1976; SRDB Ref ID: 60326 

ZPR-III, 2010, ZPR-3 Assembly 12: A Cylindrical Assembly of Highly Enriched Uranium, Depleted 

Uranium and Graphite with an Average U-235 Enrichment of 21 Atom%, IEU-COMP-FAST-004, R. 

M. Lell; Argonne National Laboratory; September 30, 2010; SRDB Ref ID: 118561, PDF p. 9 
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Attachment 1: Data Capture Synopsis

Table A1-1: Data Capture Synopsis for Argonne National Laboratory-West

Data Capture Information General Description of Documents Captured Date 

Completed

Uploaded 

to SRDB

Primary Site/Company Name: Argonne National 

Laboratory-West (ANL-W) 

DOE 1949-2005 

Physical Size of the Site: Approximately 259 acres with 52 

major buildings. 

Site Population: In 1993 there were approximately 860 

employees on site. 

ANL-W facilities include Experimental Breeder Reactor 

(EBR) I and II, Zero Power Reactor (ZPR), Zero Power 

Physics Reactor (ZPPR), the Boiling Water Reactor 

Experiment Reactors (BORAX), Hot Fuel Examination 

Facility (HFEF), Transient Reactor Test Facility (TREAT), 

Argonne Fast Source Reactor (AFSR), and the Neutron 

Radiography Facility (NRAD). 

The following documents originated at ANL-W and are, in fact, ANLW documents. However, since ANL-W was integrated into Idaho 

National Laboratory (INL) in 2005 they were identified in the 

collections at INL: Incident reports and summaries, air sample results 

including actinide concentrations, whole body count results, fecal and 

urine bioassay results, facility routine survey schedules, ZPPR neutron 

measurements and dose assignments, radiation work permits, 

radiological surveys, environmental reports, tritium contamination at 

HFEF, environmental air samples, operational, safety, and Health 

Physics monthly reports, air dust survey results, material transfers, safe 

work permits, Health Physics logs, respiratory protection, effluent 

releases, facility plot and floor plans, neutron sources, neutron 

radiography, decontamination and decommissioning of EBR-I, EBR-II 

daily radiation safety logs, plutonium reactor core loadings, 

photographs, waste management and forecasts, 1968 EBR-II Sodium 

Boiler fire, non-routine sample processing, smear sample data, cell 

entry exposure logs, temporary badge reports, alpha contamination 

survey reports, interviews with site experts, surveys of EBR-II, TREAT, 

and ZPPR, reactor safety inspections, a photo of the first plutonium 

produced at EBR, EBR external dosimetry files, contamination reports, 

operations and maintenance manuals, Reactor Development Program 

progress reports, EBR film reports, exposure summaries, EBR-II fuel 

studies, the Fuel Cycle Facility description, an addendum to the 

HFEF/N facility safety report, a documented communication, laboratory 

and office building survey logs and air samples, radiation dose histories 

for a group of 50 early ANL-W workers, accumulated personnel 

exposure summaries, 1951-1962, sample record sheets for selected 

ANL-W workers, a 1962 Burial Ground survey log sheet, and IDO 

H&S sample record sheets for EBR.

11/05/2015 2,239

State Contacted: NA Contacting the state was not considered necessary since Argonne 

National Laboratory-West, now known as the Materials and Fuels 

Complex of Idaho National Laboratory, is an active DOE site and 

cooperates with relevant data collection. 

08/03/2015 0
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246 of 260 

Data Capture Information General Description of Documents Captured Date 

Completed

Uploaded 

to SRDB

Albany Research Center (ARC) History of remediation at ARC, where surveys were performed by 

ANL.

11/01/2012 1

Argonne National Laboratory East (ANL-E) 1955 Industrial Hygiene and Safety Division monthly reports, monthly 

Operation Clean Sweep status reports, BORAX-I design calculations, 

ZPR-III operating instructions, EBR-II technical specifications, the 

1971 TREAT manual, the first plutonium fuel loading of EBR-I, 

Radiation Safety Section monthly reports, air sample data, ZPR-III 

hazards report, Health Physics reports, 1958-1961 bioassay reports, an 

EBR-II loss of water incident, a 1965 explosion at EBR-II, plutonium 

production at EBR-I, EBR-I shielding measurements, radiation dose 

histories for 11 selected workers, and radiation dose histories for a 

group of 50 early ANL-W workers.

10/23/2015 111

Atlantic Richfield ZPPR external exposure reports and study, 1967-1968. 09/12/2005 3

Battelle Memorial Institute - King Avenue A listing of DOE research papers 1949-1983, annual audits of the 

Plutonium Laboratory, and a status report on the decontamination and 

decommissioning of the Plutonium Laboratory.

01/08/2013 3

Brookhaven National Laboratory A 1974 comparison of external exposures at AEC facilities and a 1949 

description of the AEC Four Point Cancer Program.

11/18/2008 2

BWX Technologies, Inc. Documentation of ZPPR fuel element production at the NUMEC Parks 

Township Facility.

05/24/2005 1

Claimant Provided A matrix showing Argonne's part in nuclear weapon production. 04/18/2005 1

Curtiss-Wright, Cheswick, PA Material transfer records with package and fuel pin surveys. 04/26/2009 3

Dade Moeller The Leonard Koch book describing EBR-II. 05/26/2015 1

DOE Albuquerque The 1984 hazard classification of ANL-W nonreactor nuclear facilities. 04/15/2010 1

DOE Carlsbad A report on the measurement of the 10-B(n,He) and 6-Li(n,He) reaction 

rates in ZPPR-13C core.

08/10/2010 1

DOE Germantown Material receipts at the FMPC from Argonne, the DOE Oak Ridge 

Operations Records Holding Area search request procedure, and the 

2010 Argonne EEOICPA Reasonable Search Procedures.

03/07/2011 3

DOE Legacy Management - Grand Junction Office Material transfers, February 1955 accountability station codes, early 

production progress reports showing work at Argonne, the disposition 

of CANEL equipment, and a 1959 index to reactor fuel data sheets.

08/25/2011 21

DOE Legacy Management - Morgantown An environmental, safety, and health needs site summary, June 1955 

accountability station codes, EPA proposed standards for airborne 

radionuclide releases, and recycled uranium project data.

03/03/2011 6
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247 of 260 

Data Capture Information General Description of Documents Captured Date 

Completed

Uploaded 

to SRDB

DOE Legacy Management - MoundView (Fernald 

Holdings, includes Fernald Legal Database)

Material transfers, Materials Testing Reactor (MTR) meeting notes, 

1983 semiannual index of DOE unusual occurrence reports, a 1987 

DOE precious metal report, 1982 and 1984 ANL-W effluent discharge 

reports, material transfers and a report on the incineration of solid 

wastes.

05/13/2010 13

DOE Legacy Management - Westminster 1969 and 1982 Rocky Flats reports on processing ZPPR fuel fabrication 

residues.

10/23/2014 2

DOE Oak Ridge Operations Office A report on neutron and gamma exposures from radiation accidents, 

internal audit criteria for industrial hygiene programs, a 1945 discussion 

on research and production processes, and planning for future use and 

disposition of U-233. 

07/09/2012 4

DOE Oak Ridge Operations Office - Records Holding Task 

Group (RHTG)

A monthly production report, resumes of activities, and uranium 

production statistics with a thorium inventory.

04/08/2011 4

DOE Office of Scientific and Technical Information 

(OSTI)

Surveys of irradiation facilities, performance testing of personnel 

dosimetry services, material transfers, breeder fuel development reports, 

EBR-II operational Health Physics problems, a power reactor fuel 

symposium, and a 1953 Reactor Engineering Division quarterly report. 

05/22/2015 16

DOL - Paragon Boron-10 project reports, 1952 operational reports, a requested material 

transfer, and a low-level mixed waste assessment in support of the DOE 

programmatic waste management environmental impact statement.

01/23/2012 6

Energy Technology Engineering Center (ETEC) An exposure summary for the de-cladding of an EBR-II fuel element. 12/01/2005 1

Federal Records Center (FRC) - Denver A 1967 review of criticality accidents and the inclusion of ANL-W in 

the study of the recovery of Pu-244 from Mark 18A targets.

01/31/2012 2

Federal Records Center (FRC) - Kansas City 1983-1985 nuclear fuel cycle and waste management field work 

packages.

08/15/2008 1

Federal Records Center (FRC) - San Bruno LBNL Health Physics quarterly reports, selected nuclear materials 

management and safeguard systems samples, 1960 operations 

documents, and isotopes received 1949-1960.

08/02/2012 9

General Atomics Material transfer reports 1964-1970. 11/02/2005 1

General Electric - Evendale A brief history of GE at INL. 11/12/2010 1

General Electric - Vallecitos A 1968 progress report on the recovery of plutonium from ZPPR 

wastes.

02/02/2007 1

Hagley Museum & Library Trip reports to ANL-W. 09/29/2010 3

Hanford Hanford periodic operational and radiation protection reports 

mentioning ANL-W, trip reports, disposition of EBR-II test fuel, annual 

inventories by mass and activity for 1983-1991, and test program 

correspondence.

10/21/2014 37
SEC-00224 02-18-16 ANL-W 

248 of 260 

Data Capture Information General Description of Documents Captured Date 

Completed

Uploaded 

to SRDB

Interlibrary Loan An Argonne history, neutron tomography of fuel bundles, irradiated 

fuel examination at HFEF, proceedings of the 1993 incineration 

conference, proceedings of a 1964 symposium on surface 

contamination, and the operation of the EBR-II Fuel Cycle Facility.

08/31/2015 7

Internet - Defense Technical Information Center (DTIC) A bibliography of occupational dose reduction reports, the final 

proposed environmental statement for liquid metal fast breeder reactors, 

EBR-II reports, corrosion of metal waste forms and metallic uranium, 

and the 1998 Defense Nuclear Facilities Safety Board report to 

Congress.

10/04/2013 8

Internet - DOE The DOE Standard: Guide of Good Practices for Occupational 

Radiological Protection in Plutonium Facilities.

05/11/2007 1

Internet - DOE Environmental Management Linking Legacies Chapter 3: Wastes. 10/28/2007 1

Internet - DOE Legacy Management Idaho National Laboratory decontamination and decommissioning logic 

diagrams and the April 1993 DOE interim mixed waste inventory 

report, volume 2.

08/27/2014 3

Internet - DOE Legacy Management Considered Sites The April 1993 DOE interim mixed waste inventory report, volume 1. 04/24/2012 1

Internet - DOE Oak Ridge Operations Office The Interim Action Proposed Plan for Fuel and Flush Salt Removal 

from the Molten Salt Reactor Experiment, Oak Ridge National 

Laboratory, Oak Ridge, Tennessee.

01/14/2014 1

Internet - DOE OpenNet Linking Legacies Appendix B, summaries of purchase order status 

reports, a plutonium balance report, a SNAPTRAN monitoring report, 

Argonne history, radiation protection standards, history of the 

Manhattan District Research Division, the cross section of carbon, 1951 

stable isotopes shipments and requests, 1965 Idaho Falls Weather 

Bureau Research Station Program and follow-up of survivors of serious 

radiation accidents.

09/17/2015 18

Internet - DOE OSTI Chemical Engineering Division periodic reports, EBR-1 progress report 

for 1951-1953, and the Tiger Team Assessment of INEL.

09/16/2015 13

Internet - DOE OSTI / SC&A The 1945-1955 summary of radiation accidents and incidents. 02/21/2007 1

Internet - DOE OSTI Energy Citations Fuel development reports, material transfers, examinations of irradiated 

fuel specimens, overview of the Hot Fuel Examination Facility, 

capabilities of ZPPR to support active interrogation research of special 

nuclear material, EBR-II rotating plug maintenance, and the chemical 

processing of Pu-238.

03/28/2013 23
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249 of 260 

Data Capture Information General Description of Documents Captured Date 

Completed

Uploaded 

to SRDB

Internet - DOE OSTI Information Bridge Periodic radioactive waste reports, environmental reports and plans, 

plutonium contamination monitoring in the HFEF hot cells, the sodium 

process facility, fast neutron radiography research, neutron monitoring, 

glove box usage and decontamination reports, ZPPR progress reports, 

spent fuel treatment, transuranic waste reports, fast reactor safety, safety 

analysis of the use of plutonium in ZPR-6 and -9, EBR-II operational 

and safety reports, and the confirmatory radiological survey of the 

BORAX-V turbine building.

03/29/2013 167

Internet - DOE OSTI Information Bridge / SC&A / INL Health Physics aspects of the 07/22/1954 BORAX experiment. 07/06/2007 1

Internet - DOE OSTI SciTech Connect Reactor physics, development, and engineering progress reports, 

radioactive waste management, fuels and materials periodic reports, the 

shutdown of EBR-II, Nevada National Security Site waste acceptance 

procedures, environmental reports, a hot cell window shielding analysis, 

the remote treatment facility for transuranic waste, metallurgy division 

periodic reports, a report on BORAX-V, Reactor Development Program 

progress reports, EBR-II summaries of operating experience, a nuclear 

waste semiannual progress report, failed fuel monitoring and 

surveillance techniques, and the manufacture of fuel and blanket 

elements for BORAX-IV.

11/23/2015 208

Internet - DOE NNSA Nevada Site Office No relevant documents identified. 02/13/2015 0

Internet - Energy Employees Claimant Assistance Project 

(EECAP)

Soils characterization and remediation reports. 03/24/2014 3

Internet - Google Environmental reports and plans, change notice 2 to the DOE Standard 

Guide of Good Practices for Occupational Radiological Protection in 

Plutonium Facilities, DOE occupational radiation exposure reports, the 

EBR-I designation as a national historic landmark, engineering 

evaluations, groundwater evaluations, records of decision for ANL-W 

facilities, hazards summary reports, transuranic waste reports, 2011 

plutonium contamination in ZPPR, a waste disposal inventory, material 

transfers, incident reports, radiological effluent reports, Materials 

Testing Reactor design, safety and health assessments, a history and 

status of fast breeder reactor programs, the EBR-II containment 

building, EBR-II fuels and materials, Reactor Development Program 

Progress reports, and EBR-II test and operating experience.

12/05/2015 250

Internet - Hanford DDRS Hanford periodic operational reports mentioning ANL-W, a trip report, 

production of plutonium for ZPPR, and dissolution of spent ZPPR fuel 

in nitric acid.

03/25/2015 26
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Data Capture Information General Description of Documents Captured Date 

Completed

Uploaded 

to SRDB

Internet - Hathitrust The compendium of radiation safety information for plutonium. 10/25/2012 1

Internet - Health Physics Journal Environmental iodine monitoring, the high level waste tank closure 

project, the transuranic waste absorbent addition project, and testing hot 

cell shielding in the Fuel Conditioning Facility (FCF).

05/19/2015 4

Internet - IAEA INIS EBR-II advanced driver fuels and removal of a perchlorate 

contaminated hot cell acid fume system.

02/13/2015 2

Internet - Journal of Occupational and Environmental 

Hygiene

No relevant documents identified. 02/13/2015 0

Internet - National Academies Press (NAP) A nuclear technologies timeline and electrometallurgical techniques for 

spent fuel treatment.

02/12/2015 2

Internet - National Institute for Occupational Safety and 

Health (NIOSH)

The 2004 and 2006 reports on residual radioactive and beryllium 

contamination.

01/25/2007 2

Internet - NRC Agencywide Document Access and 

Management (ADAMS) 

Spent fuel and radioactive waste inventories, waste management papers 

and reports, environmental impact statements for spent fuel and highlevel waste storage and disposition, surplus plutonium disposition, 

natural phenomena hazards design and evaluation criteria, and the 

examination of spent fuel rods after 15 years in dry storage.

12/11/2014 54

Internet - Oak Ridge National Laboratory (ORNL) Library ORNL periodic operational and research reports mentioning ANL-W, 

the effect of various materials on unfired Vycor glass, secondary 

gamma ray production from californium fission spectrum neutrons, and 

the interim assessment of the denatured U-233 fuel cycle feasibility.

04/08/2013 30

Internet - University of Hawaii, Manoa The 1965 airborne gamma survey of exposure rates in the National 

Reactor Testing Station Area. 

11/08/2007 1

Internet - University of North Texas ANL 1957 and 1958 annual reports, periodic Reactor Division reports, 

the facility for photographing meltdown experiments in TREAT, a 

hazard summary addendum for EBR-II, and the preliminary design and 

hazards report for BORAX-V.

06/25/2015 11

Internet - US Army Corps of Engineers (USACE) No relevant documents identified. 10/28/2014 0

Internet - US Environmental Protection Agency NEPIS A 1997 Idaho Superfund progress report, Records of Decision, and a 

1987 mixed energy waste study. 

12/19/2014 4

Internet - US Transuranium and Uranium Registries No relevant documents identified. 10/28/2014 0

Kansas City Plant The rework of cesium pellets and April 1991 daily operations reports 

for the Secretary of Energy.

10/20/2014 2

Karam, P. Personal Library Selected pages from the Chemistry of the Actinide Elements, Second 

Edition Volume 1.

04/11/2007 1

Lawrence Livermore National Laboratory Heavy Element Program for HUTCH and Program Need for Extended 

Intense Steady State Irradiations at the ICT Accelerator.

04/28/2015 2
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251 of 260 

Data Capture Information General Description of Documents Captured Date 

Completed

Uploaded 

to SRDB

Los Alamos National Laboratory An accelerator characterization report of DOE facilities, a request for 

Np-237, and an Np-237 project report.

08/15/2014 3

Los Alamos National Laboratory - LAHDRA A 1952 survey of AEC facilities, quantities and characteristics of 

contact-handled low-level mixed waste, a 1981 monthly operational 

Health Physics report, a history of critical experiments, and the 

environmental impact statement for storage and disposition of weaponsusable fissile materials.

12/06/2007 5

Metals and Controls Corporation Confirmation that Metals and Controls supplied fuel elements for ANLW.08/24/2004 1

Missouri Department of Natural Resources Plutonium working group reports. 10/01/2008 2

Mound Museum The radiometric determination of uranium in soil and air. 07/11/2008 1

National Archives and Records Administration (NARA) -

Atlanta

A 1947 report on AEC research and development projects, 

correspondence on UF6 to UF4 processing, volumes 1 and 2 of the 

DOE indoor radon study, and documentation pertaining to the U-233 - 

thorium project.

05/20/2008 4

National Archives and Records Administration (NARA) -

Chicago

Periodic reports, Reactor Engineering Division work programs, EBR-I 

work schedules, material transfers, the 1951 redesign of air sampling 

equipment to meet a lower plutonium maximum permissible 

concentration, personnel access lists, EBR-I shielding measurements, 

1954 Industrial Hygiene and Safety Division equipment requirements, 

schedule of EBR operations, EBR plutonium loading, and EBR-I 

operating instructions.

06/12/2015 88

National Archives and Records Administration (NARA) -

College Park

A requisition for graphite for the Materials Test Reactor and monthly 

SF material inventories and progress reports.

09/12/2013 2

National Archives and Records Administration (NARA) -

Seattle

Incident reports, facility Health Physics reports, employee thyroid 

counts, a 1966 Burial Ground report, neutron dosimetry data, 

radioactive waste reports, material shipment records, and a 

contamination survey of product bottles.

12/18/2014 30

National Institute for Occupational Safety and Health 

(NIOSH)

Annual and semiannual reports to Congress, the rolling of billets at 

Bethlehem Steel, Rocky Flats waste buried in the subsurface disposal 

area, worker outreach meeting minutes and confirmation, the 1963 

plutonium aerosol monitoring program, EBR plutonium bioassay 

results, ZPPR radiation survey reports, researcher notes, a report on 

aquifer contamination, a 1953 Chemical Engineering Division summary 

report, and a DOE classification release letter.

12/07/2015 27

National Institute for Occupational Safety and Health 

(NIOSH) / SC&A

The needs assessment for the former worker medical surveillance 

program at INEL and highly enriched uranium working group reports.

02/16/2006 6
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252 of 260 

Data Capture Information General Description of Documents Captured Date 

Completed

Uploaded 

to SRDB

Nevada Test Site A summary of the volume of ANL-W waste buried at the Nevada Test 

Site.

10/01/2003 1

Nuclear Metals, Inc. Notations regarding the fabrication of EBR-I fuel rods. 05/24/2012 1

Nuclear Materials and Equipment Corporation (NUMEC) ZPPR fuel contract documents, ZPPR fuel shipping data, periodic 

operations reports, and the 1969 report of a compliance meeting closeout.

04/07/2015 17

Nuclear Regulatory Commission Non-Public Holdings 1979 requests for authorization as shipper. 06/04/2012 2

Nuclear Regulatory Commission Public Document Room NUMEC AEC license documents pertaining to the manufacture of 

ZPPR fuel elements, the fabrication of EBR-II fuel elements, and the 

analysis of West Valley terminal waste forms including ANL waste.

12/19/2014 7

Nuclear Testing Archive (NTA) A cutaway photograph of ZPPR. 05/12/2015 1

Oak Ridge Library for Dose Reconstruction A list of AEC incidents, a material transfer from ORNL to ANL-W, and 

a report on methods of solid waste management.

04/27/2011 3

Oak Ridge National Laboratory (ORNL) Periodic ORNL reports that mention ANL-W, abstracts of reports added 

to the ORNL libraries, and dose measurements at the Bulk Shielding 

Facility.

02/07/2013 21

Ohio Department of Health An environmental and waste management report. 11/03/2008 1

ORAU Team Technical basis documents, documented communications, dose 

reconstruction procedures, the evaluation of incomplete internal 

dosimetry records, compiled internal dosimetry spreadsheets, and a 

2004 Argonne directory.

07/22/2015 57

Pantex Plant Records transfer control documents indexed from 1952-1990. 02/06/2008 1

Rocky Flats Environmental Technology Site (RFETS) Radiation exposure of Rocky Flats employees during the manufacture 

of ZPPR fuel elements.

05/15/2006 1

SAIC 1967 and 1971 radiation exposure summaries. 09/02/2004 2

Santa Susana Field Laboratory (SSFL) The proposed work plan for the examination of EBR-II sodium cold 

traps and a user authorization for the pulsed neutron calibration of 

flowmeters.

08/10/2009 2

Savannah River Site (SRS) Periodic reports mentioning ANL-W, a temporary procedure for 

receiving and packaging EBR-I fuel, thorium slug failures, and a trip 

report to the neptunium information meeting.

02/08/2012 15

S. Cohen & Associates (SC&A) Inventory and manufacturing statements, an MTR progress report, 

Chemical Processing Plant (CPP) monthly reports, and a citation for an 

evaluation of the ZPPR roof filter system.

06/13/2011 66
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Data Capture Information General Description of Documents Captured Date 

Completed

Uploaded 

to SRDB

SC&A / INL Analysis of EBR-II and ZPPR stack effluents, BORAX surveys and 

sample analyses, decontamination and decommissioning of EBR-I, 

environmental reports, studies of the ANL-W industrial waste ponds, 

groundwater reports, long-lived gross beta air concentrations at EBR-II, 

radioactive waste management reports, radio ecological effects on 

vicinity populations, Burial Ground air samples, inventory and 

manufacturing statements, CPP monthly reports, effluent analyses, 

environmental monitoring data and reports, meeting minutes, EBR-II 

fact sheet and photographs, and episodic releases.

06/24/2010 215

SC&A / Internet - DOE OpenNet The 1944-1994 history of plutonium production, acquisition, and 

utilization.

10/28/2014 1

SC&A / Internet - Google Savannah River Site 50 year report. 04/22/2008 1

SC&A / Internet - University of North Texas The BORAX V design and hazards summary report. 10/13/2014 1

SC&A / Pinellas Plant A 1993 waste generation and minimization report. 06/24/2010 1

SC&A / SSFL ZPPR fuel specifications, the de-cladding of gas tagged EBR-II fuel 

elements, draft letter proposing Atomics International as the supplier of 

ZPPR fuel plates, and the follow-up to the meeting on supplying ZPPR 

fuel plates.

06/24/2010 4

Senator John Heinz History Center A history of Westinghouse and atomic power. 12/20/2007 1

Southern Illinois University, Edwardsville, IL Transcript of day two of the Fifty-Sixth Meeting of the Advisory Board 

on Radiation and Worker Health.

10/29/2008 1

Unknown A 2000 review of criticality accidents, hazard summary of EBR-II, 

design summary of TREAT, 1958 BORAX atmospheric releases, 

analysis for the decommissioning of BORAX-V, functional and 

operational design requirements for decontamination and 

decommissioning of the EBR-I Mark-II NAK, DOE occupational 

exposure reports, dose estimation from a plutonium contaminated hand 

wound, ANL-W skin contaminations, the 1989 ANL-W routine 

bioassay program, radiation exposure summaries, proceedings of the 

United Nations 1955 Geneva conference, and occupational experience 

with BORAX and ZPR-III.

02/10/2011 73

Unknown / SC&A A 1957 report of core changes from the EBR-I meltdown and 

information for use in controlling radiation emergencies.

08/05/2003 2

Washington State University EBR-II reports and delayed neutron detection in the sodium loop safety 

facility.

05/29/2015 4

TOTAL N/A N/A 4,017
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Table A1-2: Databases Searched for Argonne National Laboratory-West

NOTE: Database search terms employed for each of the databases listed below are available in 

the Excel file called “ANL-W Rev 02, (83.13) 01-11-16” 

Database/Source Keywords/Phrases Hits Selected

Defense Technical Information Center (DTIC)

https://www.dtic.mil/

COMPLETED 04/21/2015

See Note above 286 0

DOE CEDR

https://www.orau.gov/cedr

COMPLETED 02/13/2015 

See Note above 0 0

DOE Hanford DDRS

http://reading-room.labworks.org/Catalog/Search.aspx

COMPLETED 05/13/2015

See Note above 0 0

DOE Legacy Management Considered Sites

http://www.lm.doe.gov/considered_Sites/

COMPLETED 10/29/2014

See Note above 264 0

DOE NNSA - Nevada Site Office

http://nnsa.energy.gov 

COMPLETED 02/13/2015

See Note above 2 0

DOE OpenNet

http://www.osti.gov/opennet/advancedsearch.jsp

COMPLETED 10/28/2014

See Note above 88 2

DOE OSTI Energy Citations

http://www.osti.gov/energycitations/

COMPLETED 07/12/2010

See Note above 1,081 1

DOE OSTI Information Bridge

http://www.osti.gov/bridge/advancedsearch.jsp

COMPLETED 07/12/2010

See Note above 971 1

DOE OSTI SciTech Connect

http://www.osti.gov/scitech

COMPLETED 10/28/2014

See Note above 315,622 14

Energy Employees Claimant Assistance Project (EECAP)

http://www.eecap.org

COMPLETED 02/13/2015

See Note above 0 0

Google

http://www.google.com

COMPLETED 10/29/2014

See Note above 189,587 48
SEC-00224 02-18-16 ANL-W 

255 of 260 

Database/Source Keywords/Phrases Hits Selected

HP Journal

http://journals.lww.com/health-physics/pages/default.aspx

COMPLETED 04/21/2015

See Note above 3,212 0

Journal of Occupational and Environmental Hygiene

http://www.tandfonline.com/loi/uoeh20#.VtS2f00o670

COMPLETED 02/13/2015

See Note above 14 0

National Academies Press

http://www.nap.edu/

COMPLETED 10/29/2014

See Note above 11,990 2

NEPIS

http://nepis.epa.gov/

COMPLETED 10/29/2014

See Note above 4 0

NRC ADAMS Reading Room

http://www.nrc.gov/reading-rm/adams.html#web-basedadams

COMPLETED 07/12/2015

See Note above 890 2

United States Army Corps of Engineers (USACE)

http://www.usace.army.mil/

COMPLETED 10/28/2014

See Note above 3 0

U.S. Transuranium & Uranium Registries 

http://www.ustur.wsu.edu/

COMPLETED 10/28/2014

See Note above 0 0
SEC-00224 02-18-16 ANL-W 

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SEC-00224 02-18-16 ANL-W 

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Attachment 2: Acronyms and Abbreviations

a.k.a. also known as 

ADAMS Agency-wide Documents Access and Management

AEC-IDO U.S. Atomic Energy Commission-Idaho Operations Office 

AFSR Argonne Fast Source Reactor

AL Analytical Laboratory

ALIPs annular linear-induction pumps

ANL Argonne National Laboratory 

ANL-E Argonne National Laboratory-East

ANL-PIO Argonne National Laboratory –Public Information Office

ANL-W Argonne National Laboratory–West

Board Advisory Board on Radiation and Worker Health 

BORAX Boiling Water Reactor Experiments

C Celsius

CAM continuous air monitor 

cc cubic centimeter

CEDR Comprehensive Epidemiologic Data Resource 

CFA Central Facilities Area (INL)

cfm cubic feet per minute 

Ci Curie

cm centimeter 

cpm counts per minute 

CPP Chemical Processing Plant

CRBRP Clinch River Breeder Reactor Plant

D&D decontamination and decommissioning 

DCAS Division of Compensation Analysis and Support 

DOE U.S. Department of Energy 

dpm disintegrations per minute

DTIC Defense Technical Information Center (database)

DU depleted uranium 

EBR-I Experimental Breeder Reactor-I 

EBR-II Experimental Breeder Reactor-II
SEC-00224 02-18-16 ANL-W 

258 of 260 

EDMS Electronic Document Management System

EECAP Energy Employees Claimant Assistance Project

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000 

ETR Engineering Test Reactor

FASB Fuel Assembly & Storage Building 

FASS Fixed air-sampling system

FCF Fuel Cycle Facility

FEF Fuels and Examination Facility

FFTF Fast Flux Test Facility

FSAR Final Safety Analysis Report

ft. foot 

g gram

H&S Laboratory Health and Safety Laboratory

HEPA high-efficiency particulate air (filter)

HFEF Hot Fuel Examination Facility

HFEF-N Hot Fuel Examination Facility-North

HFEF-S Hot Fuel Examination Facility-South 

HHS U.S. Department of Health and Human Services 

hr hour 

HS Health and Safety

HSD Health and Safety Division 

HSL Health Services Laboratory 

IBC Inter-Building Coffin 

ICPP Idaho Chemical Processing Plant

ICRP International Commission on Radiological Protection

ID interior diameter

IFR Integral Fast Reactor

in. inch

INEEL Idaho National Engineering and Environmental Laboratory 

INL Idaho National Laboratory 

IOO Idaho Operations Office (AEC) 

ITF Inspection & Test Facility

Kel-F Teflon-based protective coating 

keV kiloelectron-volt, 1,000 electron-volts 
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259 of 260 

kg kilogram 

kW kilowatt

L&O Building Laboratory & Operations Building 

lb pound 

LMFBR Liquid Metal Fast Breeder Reactor

LMR Liquid Metal Reactor

LOD Limits of detection

LOF loss of flow 

M meter

MED Manhattan Engineer District

MeV megavolt-electron, 1 million electron-volts 

MFC Materials and Fuels Complex (new name for ANL-W)

MFP/MAP mixed fission products/mixed activation products 

min minute

MOX mixed oxide (fuel)

MPC maximum permissible concentration

mR milliroentgen

mr/hr milliroentgen/hour

mrem millirem

MSD Materials Science Division

MTR Materials Test Reactor

MW megawatt

MWe Megawatt electric (power plant output) 

n/cm

2

/sec neutrons per cubic centimeter per second 

NAD Nuclear Accident Dosimeter

NARA National Archives and Records Administration 

NDA Laboratory non-destructive analysis

NIOSH National Institute for Occupational Safety and Health

NOCTS NIOSH OCAS Claims Tracking System

N-RAD Neutron Radiography Facility 

NRAD Neutron Radiography Facility 

NRC Nuclear Regulatory Commission

NRTS National Reactor Testing Station

NTA Film Nuclear Track A film
SEC-00224 02-18-16 ANL-W 

260 of 260 

OCAS Office of Compensation Analysis and Support 

OD outer diameter

ORAU Oak Ridge Associated Universities

ORAUT Oak Ridge Associated Universities Team

ORT operational reliability testing

OSTI Office of Scientific and Technical Information

OTIB ORAUT Technical Information Bulletin

PBF Power Burst Facility

PDF portable document format 

PRNC Puerto Rico Nuclear Center

QA quality assurance

RAS Reactor Analysis and Safety Division

RAS-TREAT Reactor Analysis and Safety Division-Transient Reactor and Experiment Test

RBCB run-beyond-cladding-breach

RESL Radiological and Environmental Sciences Laboratory 

RSWF Radioactive Scrap and Waste Facility

RWMC Radioactive Waste Management Complex

SEC Special Exposure Cohort 

SEFOR Southwest Experimental Fast Oxide Reactor

SPM Special Material

SRDB Site Research Database

TBD Technical Basis Document

TLD thermoluminescent dosimeter

TOP transient overpower 

TREAT Facility Transient Reactor Experiment and Test Facility

TRIGA® Training, Research, Isotopes, General Atomics 

μCi microcurie

μCi/cc microcurie

w/o without 

WBC whole-body counting 

ZPR Zero Power Reactor